Enriched UO2 Research Articles

Core physics calculation study of miniature lead-bismuth cooled nuclear reactor

Aiming at the application environment of small marine nuclear power platform, a 1MW Lead-Bismuth cooled nuclear reactor design scheme was proposed. The Monte Carlo method was used to calculate the reactor core physical parameters. 469 core fuel rods were arranged in 13 turns and 90% enriched UO2 fuel loading meet the power operation of nearly 10 years, and the core size meets the environmental requirements. The calculation results of fuel consumption show that after each loading, the reactor can run continuously for 1 year at 1 MW power. The radial density and axial flux density distribution and power distribution gradient trend of the core are reasonable, indicating that the design scheme can meet the requirements of small ships. Nuclear power platform use requirements.

Prismatic-core advanced high temperature reactor and thermal energy storage coupled system – A preliminary design

Abstract This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MWth. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO2 TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

Simulation of Mechanical Behavior of Fuel Pellets With Different Geometries

Typical Pressurized Water Reactors (PWR) fuel rods are manufactured using zirconium-based alloys as cladding and slightly enriched UO2 sintered pellets as fuel. However, in the last years efforts have been made to develop Accident Tolerant Fuels (ATF) focusing mainly in new materials to replace the cladding in order to avoid the exothermic reaction with steam experienced by zirconium-based alloys under accident conditions as observed during the Fukushima Daiichi accident. In this sense, iron-based alloys appear as a possibility to replace conventional zirconium-based alloys, and the effect of the pellet geometry in the performance of iron-based alloys fuel rods shall be investigated. The fuel pellet geometry experiences changes due to irradiation can promote early gap closure, mechanical loadings to the cladding and/or bamboo effects due to the combination of loads and irradiation creep, and all these effects depend also on the cladding properties. The objective of this paper was to address the influence of geometric parameters in the fuel pellet behavior of a stainless steel fuel rod by means of structural mechanical analysis using the well-known ANSYS software. The parameters evaluated in this paper considered fuel pellet with and without chamfer and dish. The data related to the fuel pellet performance under irradiation were obtained using a modified version of the FRAPCON code considering stainless steel as cladding. Results obtained from mechanical evaluation considering the effects through the responses of the axial, radial, plastic deformations, and resulting tensions were evaluated.

Depletion Analysis of the Nigerian Research Reactor Fuel with 19.75% Enriched UO2 Material

The depletion analysis of the Nigerian research reactor fuel with 19.75% enriched UO2 have been performed using the VENTURE PC code. The matrix exponential method was selected in this work to perform the depletion analysis. The volume fraction of the materials in this mixture was calculated and multiplied by their respective atom densities to obtain the effective atom density of the nuclide in the water, Aluminium mix region of the fuel cell model. The plot of the variation of k infinity versus hydrogen to Uranium ratio was generated using Matlab programming language for processing of the computer code result. This shows that as the ratio of hydrogen to uranium in the core of the reactor is increased, the reactivity also increases by gradually increasing the fuel cell radii till it gets to the peak of 0.6193. Any further increment in the radius of the fuel cell radii, the reactivity of the reactor decreases as the hydrogen to uranium ratio increases.

Comparative chemical and structural analyses of two uranium dioxide fuel pellets

A comparative investigation of the trace element, U and Pb isotopic compositions, and structural attributes of two uranium dioxide fuel pellets was conducted for nuclear forensic applications. One fuel pellet consists of natural UO2 (0.71% 235U), while the other contains enriched UO2 (3.98% 235U). Numerous impurities, including transition metals, lanthanides, and main group elements were detected in both fuel pellets with total trace element contents that vary between 9.77 and 11.37 μg/g. The relative abundances of the transition metal impurities in the natural UO2 fuel pellet are linked to the source U ore and provide important insights into its provenance. X-ray fluorescence spectroscopy was utilized to investigate and compare the macro-(cm)-scale distribution of impurities within each of the fuel pellets. In addition, structural properties of the fuel pellets were examined using powder X-ray diffraction (PXRD) and Raman and infrared spectroscopy. Differences in the intensity of the T2g U-O stretching vibrational mode were observed in the Raman spectra of the fuel pellets. Pb isotope ratios were investigated to potentially differentiate between the natural and enriched UO2 samples and are linked to the process history of these materials.

Preliminary control strategies of megawatt-class gas-cooled space nuclear reactor with different control rod configurations

A megawatt-class, gas-cooled fast reactor was designed to provide 2.2MWth power for space application. The reactor concept adopts 90% enriched annular UO2 rods as the fuel and He-Xe mixture as the coolant. The thermal power is converted into electric power through Brayton turbines. The neutron flux distributions of four selected reactor configurations with different control rod arrangement are calculated using Monte Carlo code. In addition, calculations with different combinations of inserted control rods are conducted to investigate their effects on the power profile. Results show that control rods arranged in the central region help reduce the neutron flux peak and the radial flux peaking factor. The outer control rods, however, show limited effects on the central power profile. Instead, they only affect the region of the control rod neighborhood. Besides, since the outer control rods will increase the radial peaking factor, they are better to be used as safety rods for emergency. Based on the calculation and analysis, three control strategy candidates are proposed considering power profile flattening and control flexibility.

Evaluation of Non-Oxide Fuel for Fission-Based Nuclear Reactors on Spacecraft

This project investigated the use of uranium nitride (UN) and uranium carbide (UC) reactor fuel and compared their performance to uranium oxide (UO2) in a nuclear reactor for space-based applications. As a baseline for analysis, the Prometheus Project reference reactor module was considered: a gas-cooled fast reactor using highly enriched UO2 fuel with 1 MW of thermal power output and a 15-year core life. An estimate of the temperature feedback effect on reactivity was made for each fuel type at the beginning, middle, and end of core life; results for each fuel were compared. This analysis indicates that UN-fueled reactors may exhibit a stabilizing negative reactivity feedback for increasing temperatures and that this benefit persists in the face of fuel composition changes over core life. The benefit of increased uranium loading density was assessed through a quantitative estimate of overall core weight for each fuel. It was found that weight savings on the order of 1000 kg can be realized for a reactor of this size by using either UC or UN rather than UO2.

Using Thorium Based Fuel in AP1000: Steady State Analysis

The feasibility of utilizing Thorium-Plutonium Mixed Oxide in the Westinghouse AP1000 Advanced Passive pressurized water reactor is examined under steady-state, beginning of life conditions. Initial core loading of the reactor has three types of UO2 fuel assemblies with the U235 enrichments as: 2.35w/o, 3.40w/o and 4.45w/o. In this paper, one-third of the assemblies are replaced by (Th-Pu)O2 fuel in two arrangements; the first with a blanket of (Th-Pu)O2 fuel replacing the 4.45% enriched UO2 assemblies surrounding the low enriched UO2, and in the second some of the UO2 assemblies are replaced in a way creating a ring of (Th-Pu)O2 fuel in the core. The reactor is modeled using QUARK code. The required cross-section data for QUARK have been generated using WIMSD5. The results of the steady state analysis show that introducing (Th-Pu)O2 fuel into AP1000 would not negatively impact the reactor's safety as the criteria mentioned in design control document are met. For (Th-Pu)O2 blanket and ring loading arrangements, the power peaking factor is less or equal the design limit. Over the length of the hot channel, the MDNBR varies with a minimum above the design limit for blanket and ring loading arrangements. This work provides the basis of the study on fuel behavior and thermal hydraulic analysis of AP1000 to evaluate the safety aspect of various core loading patterns under anticipated and accidental conditions.

Validation of the fission yield and decay data libraries with the 10 s-delayed [formula omitted] fission [formula omitted]-ray energy spectrum

We have measured with a LaCl3 detector the γ-ray spectrum emitted by a 235U enriched UO2 fuel rod 10 s after being irradiated with thermal neutrons. The experimental results are compared with simulations performed with the fission product yield and radioactive decay data libraries present in the most recent releases of ENDF/B, JEFF and JENDL.

Space and Time Distribution of Pu Isotopes inside The First Experimental Fuel Pin Designed for PWR and Manufactured in Indonesia

The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared and planned to be tested in power ramp irradiation. An irradiation test should be designed to allow an experiment can be performed safely and giving maximum results of many performance aspects of design and manufacturing. Performance analysis to the fuel specimen shows that the specimen is not match to be used for power ramp testing. Enlargement by 0.20 mm of pellet diameter has been proposed. The present work is evaluation of modified design for important aspect of isotopic Pu distribution during irradiation test, because generated Pu isotopes in natural UO2 fuel, contribute more power relative to the contribution by enriched UO2 fuel. The axial profile of neutrons flux have been chosen from both experimental measurement and model calculation. The parameters of ramp power has been obtained from statistical experiment data. A simplified and typical base-load commercial PHWR profile of LHR history has been chosen, to determine the minimum irradiation time before ramp test can be performed. The data design and Mat pro XI materials properties models have been chosen. The axial profile of neutrons flux has been accommodated by 5 slices of discrete pin. The Pu distribution of slice-4 with highest power rate has been chosen to be evaluated. The radial discretion of pellet and cladding and numerical parameter have been used the default best practice of TU. The results shows that Pu 239 increased rapidly. The maximum burn up of slice 4 at upper the median slice, it reached nearly 90% of maximum value at about 6000 h with peak of 0.8%a Pu/HM at 22000 h, which is higher than initial U 235. Each 240, 241 and 240 Pu grows slower and ends up to 0.4, 0.2 and 0.18 % respectively. This results can be used for verification of other aspect of fuel behavior in the modeling results and also can be used as guide and comparison to the future post irradiation examination for Pu isotopes distribution.

Analysis of gamma ray dose for dried up pond storing low enriched UO2 fuel

Gamma ray dose is calculated for loss of coolant accident in spent fuel pond (SFP) storing irradiated fuels used in light water reactors. Influence of modelling of fuel assemblies, source distributions, and loading fraction of fuel assemblies in the fuel rack on the dose are investigated.

Study of criticality safety and neutronic parameters of UO2 fuel in MTR research reactors using the MCNP4C code

The low enriched uranium UO2 fuel performance in MTR-22MW reactor is investigated in this paper with One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) to increase the scientific utilization of the reactor. The MCNP4C code was used to estimate the criticality safety and the neutronic parameters of the MTR-22MW before and after replacing the U3O8-Al original fuel by the UO2 one. The re-evaluated criticality safety parameters and the new neutronic characteristics of this reactor showed that the replacement of U3O8-Al Fuel Elements (FEs) by UO2 FEs would not significantly affect the neutronic characteristics of the MTR-22MW as well as its criticality for both the ONT and the TNTs. This result leads to say that the replacement of U3O8-Al FEs by UO2 FEs will not affect the safety of the reactor.

Analysis of Burnable Poison Effect on Combined (Th-U)O<sub>2</sub> Fuel Cycle Performance in 800 MWt Long-Life PWR

Thorium based fuel especially on thermal reactor has been a great future sight. To produce electricity, thorium should be combined by any fissile material. Many research has been conducted to examine feasibility of thorium fuel. In this present study, burnable poison feasibility on combined thorium-uranium based fuel for long life PWR at thermal power output 800 MWt has been conducted. Fuel of this reactor design has implemented ThO 2 combined by 40 – 60% enriched UO2 in an acceptable enrichment (13-20 % U-235). To reach long life operation time, fuel volume fraction examination varied for 40-60 %. Burnable poison on this design has been evaluated for 0.005-0.06 % Gd 2 O 3 , 0.5-2 % Pa-231, and 0.5-2 % Np- 237. Gd 2 O 3 has very high thermal neutron absorption cross section, small amount of relatively sufficient to reduce excess reactivity, but the effect of this material only work in early burnup step. Np-237 has 190 barns thermal neutron capture cross section which can be used to minimize excess reactivity. Pa-231 has higher thermal neutron capture cross section (280 barns) rather than Np-237 (190 barns), hence the effect of Pa-231 stronger than Np-237 for minimize excess reactivity. In the other side, Pa-231 could compensate fissile isotope in the further burnup step. The optimized design of this study reached 10 years operation time with excess reactivity smaller than 2.5% dk/k.

Comparative studies on plutonium utilization in HTTR with helium and carbon dioxide gas coolants

A high temperature engineering test reactor (HTTR) has been developed and constructed by Japan Atomic Energy Research Institute (JAERI) since 1990s. The standard HTTR has been designed for helium coolant with thermal energy output of 30 MW. HTTR uses graphite as the moderator with outlet coolant temperature of 950 °C. Instead of the pebble-bed type of fuel, the prismatic block fuel type with low enriched UO2 has been employed in this reactor.In this study we have performed comparative evaluations on plutonium oxide utilization in uranium fuel cycle of HTTR with helium gas and carbon dioxide gas coolants. The cell-burnup calculations have been performed by using a SRAC 2006 code, with the nuclear data library was derived from JENDL3.3. The 3-dimensional calculations were performed by employing a CITATION code. The results show that the reactor can achieve its criticality condition if the plutonium concentration in the loaded fuel is 5.0% or higher for both coolant materials.

Design of a PWR for long cycle and direct recycling of spent fuel

In a previous work, a new design was proposed for the Pressurized Water Reactor (PWR) fuel assembly for direct use of the PWR spent fuel without processing. The proposed assembly has four zircaloy-4 tubes contains a number of 61-element CANDU fuel bundles (8 bundles per tube) stacked end to end. The space between the tubes contains 44 lower enriched UO2 fuel rods and 12 guide tubes. In this paper, this assembly is used to build a single batch loading 36-month PWR and the spent CANDU bundles are recycled in the on power refueling CANDU reactors. The Advanced PWR (APWR) is considered as a reference design. The average enrichment in the core is 4.76%w U-235. IFBA and Gd2O3 as burnable poisons are used for controlling the excess reactivity and to flatten the power distribution.The calculations using MCNPX showed that the PWR will discharge the fuel with average burnup of 31.8MWd/kgU after 1000 effective full power days. Assuming a 95 days plant outage, 36 calendar months can be achieved with a capacity factor of 91.3%. Good power distribution in the core is obtained during the cycle and the required critical boron concentration is less than 1750ppm. Recycling of the discharged CANDU fuel bundles that represents 85% of the fuel in the assembly, in CANDU-6 or in 700MWe Advanced CANDU Reactor (ACR-700), an additional burnup of about 31 or 26MWd/kgU burnup can be achieved, respectively. Averaging the fuel burnup on the all fuel in the PWR assembly, 58 or 54MWd/kgU burnup of the fuel can be achieved, respectively. Comparing this fuel cycle strategy with that of the advanced pressurized water reactors such as AP-1000 and EPR, we find that almost the same burnup can be achieved with longer core cycle: 36 months versus 18 months for these reactors.Recycling these fuel bundles in CANDU-6 reactors requires minimizing the lattice pitch to 22cm. In order to keep the coolant void reactivity of the ACR-700 slightly negative, the calandria tube should be increased from 7.8cm to 8.2cm to decrease the moderator to fuel volume ratio.

Direct reuse of spent nuclear fuel

In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO2 fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

Study of Criticality Safety and Neutronic Performance for a 348-Fuel-Pin Ghana Research Reactor-1 LEU Core Using MCNP Code

The National Nuclear Research Institute of the Ghana Atomic Energy Commission is undertaking steps to convert the Ghana Research Reactor-1 from HEU Core to LEU. The proposed LEU core consists of 12.5% enriched UO2 fuel elements clad in Zircaloy-4 alloy. This is done in collaboration with Reduced Enrichment for Research and Test Reactor. The versatile MCNP code was used to analyse the neutronics parameters given in the SAR of HEU core, thereby characterizing the core. Subsequently, the LEU core was indentified with necessary changes to the HEU MCNP model. It was ascertained that the reactivity for the LEU core with the same number of fuel pins as the HEU was inadequate, hence the fuel pins were increased from 344 to 348. The neutron flux at the irradiation sites was found to be below the nominal value at full power for the LEU and hence the nominal power was increased to 34 kW for a nominal flux value of 1 × 1012 n/cm2.s. The parameters investigated for the HEU and LEU are shown in this paper.

Testing of HTR UO2 TRISO fuels in AVR and in material test reactors

The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance.In the 1980s, another program direction change was made to a low enriched UO2 TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions.Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO2 TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications.Fuel elements were manufactured in production scale facilities that contained near defect free UO2 TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600°C.

Design Study to Increase Plutonium Conversion Ratio of HC-FLWR Core

The innovative water reactor for flexible fuel cycle (FLWR) is an advanced reactor concept based on the well-developed light water reactor (LWR) technology. It is to be introduced in two stages to achieve effective and flexible utilization of the uranium and plutonium resources. In the first stage, the high-conversion-type reactor concept (HC-FLWR) is to be introduced, with a core that achieves a fissile Pu conversion ratio of 0.84. Then, in the second stage, the reduced-moderation water reactor (RMWR) concept can be introduced, with a breeder-type core that achieves a fissile Pu conversion ratio of 1.05. From the viewpoint of effective introduction of the high-conversion-type reactor, such as the introduction capacity of the reactor, HC-FLWR is required to further raise the fissile Pu conversion ratio to [approximately]0.95.This study aims to develop a new core design concept for the high-conversion-type core, HC-FLWR+, to achieve the higher fissile Pu conversion ratio of [approximately]0.95 under the framework of UO2 and U-Pu mixed-oxide (MOX) fuel technologies for LWRs. For raising the fissile Pu conversion ratio and controlling the void reactivity characteristics of the core, the concept of FLWR/MIX fuel assembly, which uses MOX and enriched UO2 fuel rods, is utilized.The relationships between the main design parameters and the core performance index parameters are clarified in this study. When the fuel rod diameter and the clearance range from 1.23 to 1.28 cm and 0.25 to 0.20 cm, respectively, under the same pitch of 1.48 cm, the fissile Pu conversion ratio and the core average discharge burnup range from 0.89 to 0.94 and 53 to 49 GWd/tonne, respectively (the fissile Pu conversion ratio and the burnup are subject to a trade-off). Furthermore, when 235U enrichment in the UO2 fuel rods is increased from 4.9 to 6 wt%, the fissile Pu conversion ratio improves to 0.97.From these relationships, two representative core designs with fissile Pu conversion ratios of 0.91 and 0.94 and one optional design with a ratio of 0.97 were obtained. Hence, the flexibility of HC-FLWR+ concept to achieve a higher fissile Pu conversion ratio of [approximately]0.95 has been revealed. Together with the standard HC-FLWR design, the concept covers a wide range of needs on fissile Pu conversion ratio from 0.84 up to 0.97, with design variations that are expected to be within the scope of current boiling water reactor and MOX fuel technologies.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20wt%UO2, U–25wt%UO2 and U–30wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO2 compositions.