Burnup Values Research Articles

Investigation of the hydrogen production of a laser FUSION driver thorium breeder using various coolants

In this present work, hydrogen production and neutronic calculations of a Laser Inertial Confinement Fusion Fission Energy (LIFE) driven thorium breeder using various coolants have been investigated. In the neutronic calculations for fusion driver power of 500 MWth has been examined with MCNP code. The 95 vol% Flibe or natural lithium 5 vol% TRISO coated ThC fuels have used in the neutronic calculations. Tritium breeding ratio (TBR) has been calculated as 1.08 and 1.19, respectively, for Flibe and natural lithium coolants. The energy multiplication values have been computed as 3.17 and 1.62, respectively, for these coolants. The burnup values with flibe and natural lithium have been obtained as 6 GWd/tM and 22 GWd/tM over 11 and 23 years, respectively. Also, the hydrogen production of a laser fusion driver thorium breeder using steam methane reforming (SMR), high temperature electrolysis (HTE) and sulfur-iodine (S–I) thermochemical water splitting processes have been performed. The highest hydrogen production values with flibe coolant of SMR method have been obtained as ~200 kg/s over 11 years.

Optimization of Swiss used nuclear fuel canister for final repository: Homogeneous vs. mixed loading

In this paper, the possibilities and advantages of loading mixed used nuclear fuel (UNF) assemblies into canisters for final repository are presented from a criticality aspect within the burnup credit approach (BUC). UNF coming from a Swiss reactor are taken into account at the pin-by-pin level and two canister loading patterns are studied: mixed and homogeneous (i.e. mixing UNF with various burnup values and enrichments, or not). The advantages of the mixed cases are presented in terms of criticality, together with the impact of assembly axial and radial rotations. The main outcome of the mixed loading is that a variety of UNF assemblies can be safely loaded together, including those which were previously not allowed in the homogeneous case.

Semianalytical Solution to Near-Field Temperature in Nuclear Waste Repository

The temperature field in a nuclear waste repository is an important issue with regard to the design and safety assessment of the repository. In this paper, a double-layer model for simulating the heat conduction near a single waste canister is established, and then, by applying the Laplace transform to the governing equations of the heat conduction in the buffer layer and the surrounding rock, the solutions of the temperature field are obtained in the Laplace domain. The temperature distribution near the nuclear waste canister is presented by numerical inversion of the solutions using Crump’s method in the time-space domain. Finally, the effects of parameters on the temperature on the canister surface are analyzed. The results show that the double-layer model of the heat conduction increases the maximum temperature on the canister surface by about 11.87°C compared to the single-layer model. The double-layer model is verified to be reliable by comparing with a line heat source model that has been verified by numerical calculations. The temperature on the canister surface is significantly affected by the burnup value and cooling time of the nuclear waste, the thickness of the buffer layer, the thermal conductivities of the buffer material and the surrounding rock, and the external boundary condition.

Coupled Monte Carlo-burnup and CFD analysis of coated UN and UC fuels in an HPLWR fuel assembly

Ceramic nuclear fuels may have potential benefits to be considered as nuclear fuels in future Supercritical Water Reactors (SCWR), they offer the advantage of high thermal conductivity and high-density values compared to uranium dioxide. In recent years, uranium nitride (UN) and uranium carbide (UC) are being considered as viable options for alternative nuclear fuels in SCWR, but an important issue is related to their chemical reactivity with water and nickel, which forces the use of coatings for fuel pellets, reducing heavy metal volume and reducing their thermal and neutronic performance. The main purpose of the study is the analysis of the burnup dependent performance of coated ceramic fuels in the SCWR compared with conventional UO2 fuel. Particularly, it focuses on the study of the fuel centerline temperature and the cladding surface temperature. Furthermore, the evolution of the neutron multiplication factor, the fissile inventory ratios and the main heavy isotopes composition along burnup are analyzed and compared to asses proliferation resistance issues. The use of UN fuel coated with Zirconium Carbide (ZrC) and UC fuel coated with Titanium Nitride (TiN) in the fuel assembly of the High-Performance Light Water Reactor is analyzed, and the results are compared with those of the conventional UO2 fuel. The comparative study was conducted for a hypothetic operation time of 540 days. Due to the large coolant density variation along with the active height of the fuel assembly, coupled calculations are performed using the Monte Carlo MCNP6 code and the CFD code ANSYS-CFX 19, taking into account the capabilities of CFD codes to describe the heat transfer mechanisms at supercritical conditions. The main results show significantly lower values of the centerline temperature distribution in UN_ZrC and UC_TiN fuels compared to UO2 fuel. The fuel burnup values obtained at the end of the 540 days are practically the same, a slightly higher burnup is obtained for UC_TiN fuel because of its lower amount of heavy metals in the initial composition. The evolution of the fissile inventory is practically the same for all the analyzed fuel options, around 65% of fissile inventory is still present in the core at EoC. Total plutonium production shows a decrease of 1.85% (UN–ZrC) and 4.53% (UC–TiN) with respect to UO2 fuel. Coated ceramic fuels show a slightly improved proliferation resistance compared with UO2 fuel, since a slightly smaller fraction of 239Pu is obtained in their compositions.

Neutronic Study of Fuel Depletion for the MNSR Research Reactor Using DRAGON5 Code

The aim of this work is to study the fuel depletion of 30 kW miniature neutron source reactor (MNSR). Under operating conditions of two hours per day for five days a week at a peak thermal neutron flux of 1.0E+12 n/cm2 s, the estimated core life is 10 years (200 days). The DRAGON5 code is utilizing to generate the fuel group constants and the infinite multiplication factor versus the MNSR reactor operating time, at different burn-up values. The amounts of uranium burnt up and plutonium produced in the reactor core are calculated using the DRAGON5 transport lattice code. The results are in good agreement with previous studies.

Investigation of hydrogen production potential of the LASER inertial confinement fusion fission energy (LIFE) engine

The main purpose of this paper is to research of the time-dependent hydrogen production possibility of the Laser Inertial Confinement Fusion Fission Energy (LIFE) engine using steam methane reforming (SMR), sulfur–iodine (S–I) thermochemical water splitting and high temperature electrolysis (HTE) cycles. The fuel zone of the LIFE engine contains 90 vol% Flibe coolant and 10 vol% TRISO coated minor actinides. Firstly, to examine the time-dependent hydrogen generation potential of this reactor, the time-dependent neutronic performance of the reactor has been performed with using the MCNP neutron transport code. As a result of neutronic calculations, tritium production (TBR), energy multiplication factor (M) and fuel burnup (BU) values have been computed. Secondly, the total power required for SMR, S–I and HTE cycle process have been calculated by using the time dependent energy multiplication factor obtained from the neutronic results. Depending on this total power, the change of hydrogen production with operation time in the LIFE engine has been investigated. As a result of the calculations, it has been found that the LIFE engine has a good both neutronic performance and hydrogen production by SMR, S–I and HTE cycle process.

Burnup Calculation Study of Pebble Bed Equilibrium Core

Reaktor Daya Eksperimental (RDE) is a high temperature gas-cooled reactor (HTGR) producing a 10 MW thermal with a pebble bed fuels which being developed by BATAN. The purpose of this paper is to study the burnup distribution and characteristic in the equilibrium core for different multipass recirculation method. Understanding the reactor physics, in particular the burnup, is important for optimum design and safety analysis of nuclear reactor. Related with the design approval phase of RDE, this study provides a design data which needed for the RDE’s safety analysis report, e.g. fuel pebble performance analysis. Analysis in this study was performed using PEBBED diffusion code. PEBBED is designed to solve the neutronics and thermalhydraulics parameter cases for high temperature pebble bed reactors for different fuel recirculation including once-through-then-out (OTTO) and multipass scheme. The reactor core calculation was performed by applying multipass scheme with variation from 5 passes up to 15 passes. With energy and power as inputs, the calculations produce the burnup fraction at the end of the cycle and the fast and thermal neutron fluxes in 8 energy groups. The calculation results showed that the lowest 5 passes fuel recirculation pattern has the highest and lowest minimum discharged burnup value. This is related to the average power distribution in the core which means more passes will flatter the burnup distribution including the power distribution and also seen in the value of flux and power peaking factor produced. power peaking factor produced.

Impact assessment of alternative back-end fuel cycles on geological disposal of resultant spent fuels and high level wastes

The aim of this study is to assess the impact of closing the nuclear fuel cycle on geological disposal of resultant spent fuels and high level wastes. Once-through and closed fuel cycles with different back-end scenarios are compared with regard to geological disposal. The comparison is carried out in terms of two significant parameters for permanent geological repository: disposal densities and long term radiotoxicity indices of wastes generated from fuel cycles. In the first part of the study, fuel cycles are identified and waste characteristics (amount, composition and thermal output) for the burnup values of 33,000, 40,000 and 50,000 MWd/t are estimated by using MONTEBURNS code. Then, areas needed for waste types under consideration are determined by thermal analysis carried out in ANSYS code for a reference repository concept. The second part of the study consists of an assessment of the radiotoxicities (namely ingestion-toxicity indices) of the wastes generated from the fuel cycles considered. According to the results of the disposal density analysis the once-through cycle is the most advantageous one at low burnups. However, at burnups higher than 40,000 MWd/t, the closed cycle with the standard reprocessing is better than the once-through cycle and the other closed fuel cycles. According to the results of the radiotoxicity analysis, the closed cycle with MOX recycling is more advantageous than the once-through cycle and the other closed cycles for the whole burnup range studied.

Post-irradiation examinations of [formula omitted] composites as part of the Accident Tolerant Fuels Campaign

Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO2−SiC) and diamond (UO2-diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction of UO2 composites, which have been proposed as accident tolerant fuel candidates.Both non-destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO2-SiC pellets revealed enhanced cracking when compared to UO2 pellets irradiated under similar conditions. Instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300°C. The microstructure of the UO2-diamond was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature.The postirradiation examination results cast doubt on the suitability of these additives to improve UO2 fuel performance in a way that would lead to enhanced accident tolerance.

Determination of plutonium and uranium content and burnup using six group delayed neutrons

Abstract In this study, investigation of spent fuel was performed using six group delayed neutron parameters. Three used fuels (F1, F2, and F11) which are burnt over the years in the core of Missouri University of Science and Technology Reactor (MSTR), were investigated. F16 fresh fuel was used as plutonium free fuel element and compared with irradiated used fuels to develop burnup and Pu discrimination method. The fast fission factor of the MSTR was calculated to be 1.071 which was used for burnup calculations. Burnup values of F2 and F11 fuel elements were estimated to be 1.98 g and 2.7 g, respectively. 239Pu conversion was calculated to be 0.36 g and 0.50 g for F2 and F11 elements, respectively.

Механизмы и главные причины повреждений тепловыделяющих элементов и тепловыделяющих сборок атомных электрических станций с реакторами с водой под давлением

This article is the first part of the review of publications on the problem of increasing the service life and improving the reliability of fuel assemblies (FA) and fuel rods (FR). In the next issues of the journal, Part 2 Dynamic interactions between the coolant flow and FAs and Part 3 Prediction of vibroacoustic resonances in the reactor cores of VVER-based nuclear power of this review will be published. This review part presents the results from investigations into the causes and mechanisms of FR failures, in which radioactive materials (gaseous fission products and volatile elements, in particular, krypton, xenon, iodine, and cesium) escape from the fuel pellet into the reactor primary coolant circuit. It is shown that the nuclear fuel depressurization level is influenced by all stages of the product life cycle, including design, manufacture, and operation of FAs. In particular, during operation, FA vibrations have an effect on the nuclear fuel depressurization level. As is well-known, an FA is a hydroelastic oscillatory system consisting of two interacting subsystems: a mechanical one and a hydrodynamic one. The FA vibration is due to the dynamic processes occurring in these subsystems. It is noted that the hydrodynamic subsystem has been studied to a rather insufficient extent, which is due to complexity of describing the processes through which the random hydrodynamic loads applied to the streamlined surfaces are produced, as well as the influence of flow thermal-hydraulic and acoustic characteristics on the occurrence of self-excited oscillations. Based on the defective FA operation data, the service life of leaky fuel rods and their linear heat generation rates are estimated. The evolutions of fuel operation conditions and fuel design changes are described. The available recommendations on optimizing FA designs developed taking into account the experience gained from operation of the reactors and the results of the accomplished fretting wear investigations are mentioned. It is shown that all who manufacture and operate nuclear fuel take efforts aimed at improving its quality, because it is the key indicator characterizing the vendor of fuel and nuclear power plants in the world market. Achieving longer fuel campaigns and higher burnup values while keeping or even improving the safety level are the most important criteria for nuclear fuel quality. Currently, the project called Nuclear Fuel Zero Failure Level is in the stage of implementation and obtaining the first results of joint activities of all participants.

Large-scale energy production with thorium in a typical heavy-water reactor using high-grade plutonium as driver fuel

Thorium can be introduced into the energy vector in combination with high-grade plutonium (HG-Pu). Excellent neutron economy of heavy-water moderator allows use of mixed ThO2/HG-PuO2 fuel in heavy-water reactors with high efficiency, leading to the exploitation of large world thorium reserves and extending the availability of the nuclear energy by two orders of magnitude. In the present work, the criticality calculations have been performed with the code MCNPX 3-D geometrical modeling of a typical heavy-water reactor, where the structure of all fuel rods and bundles is represented individually. In the course of time calculations, nuclear transformation and radioactive decay of all actinide elements as well as fission products are considered. Five different fuel compositions have been selected for investigations: (1) 97% thoria (ThO2) + 3% PuO2; (2) 96% ThO2 + 4% PuO2; (3) 95% ThO2 + 5% PuO2; (4) 94% ThO2 + 6% PuO2; and (5) 92% ThO2 + 8% PuO2. The behavior of the criticality k∞ and the burn-up values of the reactor have been pursued by full power operation at 640 MWel (2180 MWth) for approximately 7 years. Time calculations have been conducted with MCNPX and CINDER codes under consideration of all nuclear transformation and radioactive decay processes on the actinide isotopes and fission fragments. As the reactor allows fuel recharging at on-power operation mode, the reactor criticality has been followed down to keff,end = ~1.05. The corresponding burn-up values and operation periods for the investigated modes are (1) 18 GWd/MT and 780 days; (2) 27 GWd/MT and 1200 days; (3) 35 GWd/MT and 1560 days; (4) 44 GWd/MT and 1940 days; and (5) 60 GWd/MT and 2640 days. Among the investigated four modes, 94% ThO2 + 6% PuO2 seems a reasonable choice under consideration of the high price of the HG-Pu as driver fuel. The mixed fuel has the potential of an extensive exploitation of thorium resources. Reactor will run with the same fuel charge for approximately 5 years and allow a fuel burn-up approximately 44 GWd/MT, comparable with conventional light-water reactors (LWRs). Plutonium component of the mixed fuel will become nonprolific after few months of plant operation through the accumulation of even isotopes. Addition of few percent natural uranium to the initial mixed fuel charge will keep the 233U component below 22% and hence at nonprolific level over the entire plant operation period. Replacement of 4% ThO2 with 4% nat-UO2 will practically not change main technical parameters of the reactor.

Research activities at GRS on fuel rod behaviour during extended dry storage

Abstract As a common part of nuclear waste management strategies, used nuclear fuel is dry stored in casks. After a sufficient time of wet cooling in fuel pools, the fuel assemblies in Germany are dry stored in dual purpose casks mostly on-site until a deep geological repository will be available in the future. In the light of the extended and preliminary not foreseen timescales of up to 100 years of dry storage, open questions arise concerning the long term fuel rod behaviour. Most of these questions are linked to safety criteria e.g. the integrity and fuel encapsulation. Due to the variety of used nuclear fuels and cladding materials in Germany and the additional use of MOX fuels and very high burn-up values of 65 GWd/tHM averaged over a fuel assembly, some of the boundary conditions are unique and specific to the German situation. The GRS has a strong research program to investigate the long term behaviour of fuel assemblies under the German specific conditions of dry storage. Some aspects are presented in the following article.

Determination of fuel burnup distribution of a research reactor based on measurements at subcritical conditions

This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor (DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of linear dependence of the reactivity on the burnup of fuel bundles and the measurements at subcritical conditions. The measurements were taken for seven selected fuel bundles in two different measuring sequences. The measured burnup values have also been compared with the calculations for verifying the method and the measurement procedure. The results obtained with the three detectors have a good agreement with each other with a discrepancy less than 1.0%. The errors of the measured burnup values are within 6%. Comparison between the calculated and measured burnup values shows that the discrepancy of the C/E ratio is within 9% compared to unity. The results indicate that the method of measurements at subcritical conditions could be well applied to determine the relative burnup distribution of the DNRR.

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification

The article presents information on the validation and verification (V&V) of the first version (V1) of the EUCLID integrated code intended for safety analysis of operating or designed liquid metal (sodium, lead, or lead–bismuth) cooled reactors under normal operation and under anticipated operational occurrences by carrying out interconnected neutronics, thermal–mechanical, and thermal–hydraulic calculations. The list of processes and phenomena that have to be modeled in the integral code for correctly describing the above-mentioned operating conditions is given. Based on this list, the most high-quality experimental data are selected for carrying out the validation. It is shown that, for sodium cooled reactors, a significant number of experiments was carried out around the world on studying individual thermal–hydraulic processes and phenomena, which made it possible to perform validation of the thermal–hydraulic module. The validation of the code—as applied to description of processes that take place in fuel rods with oxide or nitride fuel and gas gap—is carried out against the results of post-pile investigations of fuel rods irradiated in fast sodium cooled research and power-generating reactors. The obtained results opened up the possibility to determine the errors of calculating such fuel rod parameters as release of gaseous fission products from the fuel and sizes of pellet and cladding in a limited range of burnup values. To perform validation of the neutronics module as applied to calculation of such parameters as power density distribution over the core and decay heat release, a sufficient number of experiments and benchmarks were selected. The results obtained from experimental operating conditions of a BN-600 reactor and startup conditions of a BN-800 reactor made it possible to estimate how correctly the integral code performs calculations of interconnected thermal–hydraulic and neutronic processes. Only a limited set of experimental investigations is available for heavy liquid metal cooled reactors. In view of this circumstance, programs for obtaining the lacking data are developed. To estimate the quality with which the experiments are modeled by means of the EUCLID/V1 integrated code, a procedure for evaluating the errors of calculation results is developed. In accordance with this procedure, the error of calculating the parameters playing the main role in the reactor safety assessment is evaluated.

Monte Carlo analysis of LWR spent fuel transmutation in a fusion-fission hybrid reactor system

Abstract The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS2M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources.

Monte Carlo calculations of the incineration of plutonium and minor actinides of laser fusion inertial confinement fusion fission energy (LIFE) engine

In this paper, neutronic analysis in a laser fusion inertial confinement fusion fission energy (LIFE) engine fuelled plutonium and minor actinides using a MCNP codes was investigated. LIFE engine fuel zone contained 10 vol% TRISO particles and 90 vol% natural lithium coolant mixture. TRISO fuel compositions have Mod①: reactor grade plutonium (RG-Pu), Mod②: weapon grade plutonium (WG-Pu) and Mod③: minor actinides (MAs). Tritium breeding ratios (TBR) were computed as 1.52, 1.62 and 1.46 for Mod①, Mod② and Mod③, respectively. The operation period was computed as ∼21 years when the reference TBR > 1.05 for a self-sustained reactor for all investigated cases. Blanket energy multiplication values (M) were calculated as 4.18, 4.95 and 3.75 for Mod①, Mod② and Mod③, respectively. The burnup (BU) values were obtained as ∼1230, ∼1550 and ∼1060 GWd tM–1, respectively. As a result, the higher BU were provided with using TRISO particles for all cases in LIFE engine.

The U-232 production in thorium cycle

The main problem of Th-232 and U-233 mixture as a reactor fuel is that it contains the admixture of U-232. The decay chain of U-232 produces penetrating gamma rays. The concentration of U-232 as a function of burnup, different kinds of a coolant and average neutron energy are investigated. The main channels of production and destruction of U-232 are presented. The assembly that was used in this paper is a critical reactor which makes it possible to change volumes of fuel and coolant. One can obtain the minimal value of U-232 saturation concentration using Pb-208 coolant and the maximal value of the critical burnup together. The calculations were carried out using MCNP6.1/MCNPX 2.7 codes and the analytical model.

Correlation between pressure tube thickness and expected discharged burnup in CANDU: Lattice calculations

CANDU (Canada Deuterium Uranium) is a pressurized-heavy-water reactor design where the fuel bundles are oriented horizontally inside the fuel channels. Each fuel channel is comprised of concentric tubes called the pressure tube and the calandria tube. As the reactor ages, due to continuous exposure to high neutron field, temperature, and pressure, a deviation from the concentric arrangement becomes more observable. As a result of irradiation and the presence of high pressure and temperature, the pressure tube undergoes changes such as diametral creep, sag, and axial elongation. The changes in the pressure tube geometries have an unfavourable impact on the operation of the reactor where the aged reactor overall power output must be reduced in order to maintain the same safety margin as the new reactor. One of the options to overcome this challenge is to utilize thicker pressure tubes. The impact of this change from the neutron economy point-of-view is evaluated in this paper; in particular its impact on the expected discharged burnup value. From this study, it is concluded that a percent increase in the thickness of the pressure tube leads to approximately one-half-of-a-percent decrease in the expected discharged burnup value.