Total Burnup Research Articles

Assessing the benefit of thorium fuel in a once through molten salt reactor

This study comprehensively evaluates thorium-based fuel utilisation in the Molten Salt Reactor (MSR) with a once-through fuel cycle. It examines the whether the use of thorium is beneficial in extending the fuel cycle length of an MSR, fuel utilisation efficiency, reactor safety, and reduction of long-lived radioactive waste. To observe those parameters, this research compares the use of thorium and uranium as fertile fuel in a single-fluid, once through MSR using Uranium-233 (U-233), Uranium-235 (U-235), and Reactor Grade Plutonium (RGPu) as well as the fissile driver. MCNP radiation transport code with ENDF/B-VII.0 continuous neutron library was used to analyse the neutronic parameters and fuel burnup in a multi-refuelling scheme for a total burnup time of 8 years. The findings indicate that the utilisation of thorium in conjunction with U-233 and U-235 significantly enhances fuel consumption and improve reactor safety. Such benefits, however, did not appear in RGPu. Moreover, thorium effectively reduces the production of long-lived radioactive waste, a crucial issue in nuclear waste management. Considering technical and environmental aspects, this study offers novel insights into using thorium as a more sustainable nuclear alternative under a once through fuel cycle, within a condition of multi-refuelling cycle which is only possible in an MSR.

Cladding of nuclear fuel with iron based alloy: Penalty and solution

Replacing zirconium-based alloys fuel cladding by that of iron in nuclear fuel of Light Water Reactors (LWRs) needs increasing the fuel enrichment to achieve the same core cycle length. The penalty of using iron based alloys fuel cladding in LWRs can be overcome by direct recycling of nuclear spent fuel in CANDU reactors. The problems associated with using zirconium-based alloys fuel cladding in LWRs is not significant in the case of CANDU reactors. This is because CANDU reactors discharge the spent fuel with low burnup which is in the withstand tolerance of zirconium-based alloys and the moderator acts as an effective heat sink in case of severe accident.The case studied in this research is direct recycling of the spent fuel discharged from the European Advanced Pressurized Water Reactor (EU-APWR) in CANDU-6 reactor. The calculations using MCNPX computational code showed that replacing fuel clad from zircaloy-4 to stainless steel-348 requires increasing the fuel enrichment from 4.95 %w U-235 to 6.8 %w U-235 with an increase of 37%. After direct spent fuel recycling in CANDU-6, the total burnup of the fuels cladded with zircaloy-4 and with stainless steel-348 are about 70 and 96 GWd/t, respectively with an increase of about 37% in the case of the fuel cladded with stainless steel which compensates the increase of the enrichment.

Implementation of in-rod axially heterogeneous thorium-uranium fuel in a typical PWR

Given the potential of the thorium fuel cycle and the predominance of PWRs in today’s commercial fleet of nuclear power plants, further investigations on the introduction of thorium-based fuel in open-cycle reactors such as PWRs are warranted. The benefits of microheterogeneous thoria-urania fuel (e.g. increased burnup / fuel cycle length, reduced radiotoxicity and increased stability of spent fuel, proliferation advantages due, feasibility of converting UO2-based fuel manufacturing plants to ThO2-UO2), are diminished by fuel manufacturing costs, the need to denature thorium sections of the fuel, thermo-hydraulic design requirements, higher fission gas release and cladding temperature gradients.This paper presents results of the neutronic performance of homogeneous and multiple heterogeneous fuel pin cell designs and explores the fundamental reasons and reaction balances between the different fuel models. The work utilises fuel models with variations in the length and composition of thorium and uranium sections. We compare the various urania-thoria fuel models with a reference PWR (North-Anna 1) with similar fissile content. We postulate a number of fuel design criteria, to appraise fuel models in trade-off studies, including improved burnup and cycle length, limitations on fuel enrichment grade and excess reactivity during burnup (typically at beginning of cycle), safety requirements such as limitations on total burnup, power peaking and feedback mechanisms. In addition, criteria are set for proliferation resistance and radiotoxicity.Verification studies were performed on fuel with homogeneous and axially heterogeneous fuel cell pins, followed by initial assembly and full-core analyses, to affirm confidence in the models employed. Our results confirm significant conversion of thorium to 233U, resulting in an improved burnup and fuel cycle length. Shut down margins and neutron economy are improved by significantly reducing beginning of cycle excess reactivity and hence reducing the need for burnable and soluble neutron absorbers. The conversion of thorium to 233U then leads to an increase in reactivity later in the cycle. To address proliferation concerns, fuel models with denatured thorium sections are introduced.We conclude that among the well-performing fuel models, fuel comprising heterogeneous pins repeating 3 cm thorium and 1 cm uranium sections, with the uranium region enriched to 20 at% 235U, and the thorium region denatured to a specific degree, performed best in terms of the fuel criteria considered. The promising results for the fuel pin cell models were confirmed with initial assembly and full core models: Burnup improvement rendered by the 3Th1U fuel models is 25.6% / 22.3% / 17.9% for the pin cell / assembly / full core models respectively, in spite of employing denaturing of thorium sections, while also fulfilling the other stated fuel criteria.

Analysis of Th and U breeding in a heat pipe cooled traveling wave reactor

Introduction: Heat pipe cooled traveling wave reactor (HPTWR) is a newly proposed heat pipe reactor. The HPTWR can achieve the low enrichment of loaded fuel, small reactivity swing, and long-term continuous operation for the power supply of decentralized electricity markets. Due to the excellent breeding capability of the HPTWR, the Th fuel is also added into the breeding fuel region of the reactor to achieve the Th-U fuel cycle in this work.Methods: The Monte Carlo code RMC is used to obtain the reactivity swing, propagation of axial power peak, burnup, and productions of bred fissile nuclides for the HPTWR with Th and U fuels.Results and Discussion: The results indicate that the HPTWR with 13.6% 235U enrichment of ignition fuel and 20% 235U enrichment of breeding fuel can continuously operate for 18.1 years without refueling when the mass fraction of 232Th in heavy metals of breeding fuel region is 33%. The propagation velocity of axial power peak and total burnup for the HPTWR with Th and U fuels is about 0.5525 cm/years and 24.72 GWd/THM during the 18.1 years operation respectively. The corresponding productions of bred 239Pu, 241Pu and 233U are about 212.99 kg of 239Pu, 0.19 kg of 241Pu and 81.58 kg of 233U at the end of cycle (EOC) respectively. The obtained results in this study demonstrate that the HPTWR can achieve the Th fuel breeding in the case of the low 235U enrichment loading (≤ 20%).

Investigation of carbon fiber–reinforced thermoplastic polymers using thermogravimetric analysis

The high chemical resistance of the advanced thermoplastic polymers, such as polyether ether ketone (PEEK), polystyrene (PS), or polyamide (PA), makes methods of the composites’ composition evaluation based on chemical etching difficult and tedious. Wider application of the composites with thermoplastic matrix creates a need to find more capable, faster, and easier method of fiber and void content determination. Thermogravimetric analysis (TGA) is known to be one of the most promising candidates. Numerous reports have shown that TGA could be successfully applied for glass-reinforced composites. In this study, carbon fiber/PEEK composite was investigated. Initial tests showed that a simple method based on total burn-up of the matrix cannot be used for such advanced composites due to the high thermal resistance of PEEK. Tests carried with various parameters and microscopic investigation of the degradation process allowed to propose an analysis method that can give a satisfactory outcome. Results obtained with carbon fiber/PEEK composite using TGA-based method matched those obtained by chemical etching. Supplementary investigation of PA-based and epoxy resin composites showed the applicability of the method to a wider group of both thermoplastics- and thermoset-based carbon fiber–reinforced composites.

Experimental evaluation of models for predicting Cherenkov light intensities from short-cooled nuclear fuel assemblies

The Digital Cherenkov Viewing Device (DCVD) is a tool used by nuclear safeguards inspectors to verify irradiated nuclear fuel assemblies in wet storage based on the recording of Cherenkov light produced by the assemblies. One type of verification involves comparing the measured light intensity from an assembly with a predicted intensity, based on assembly declarations. Crucial for such analyses is the performance of the prediction model used, and recently new modelling methods have been introduced to allow for enhanced prediction capabilities by taking the irradiation history into account, and by including the cross-talk radiation from neighbouring assemblies in the predictions. In this work, the performance of three models for Cherenkov-light intensity prediction is evaluated by applying them to a set of short-cooled PWR 17x17 assemblies for which experimental DCVD measurements and operator-declared irradiation data was available; (1) a two-parameter model, based on total burnup and cooling time, previously used by the safeguards inspectors, (2) a newly introduced gamma-spectrum-based model, which incorporates cycle-wise burnup histories, and (3) the latter gamma-spectrum-based model with the addition to account for contributions from neighbouring assemblies. The results show that the two gamma-spectrum-based models provide significantly higher precision for the measured inventory compared to the two-parameter model, lowering the standard deviation between relative measured and predicted intensities from 15.2 % to 8.1 % respectively 7.8 %. The results show some systematic differences between assemblies of different designs (produced by different manufacturers) in spite of their similar PWR 17x17 geometries, and possible ways are discussed to address such differences, which may allow for even higher prediction capabilities. Still, it is concluded that the gamma-spectrum-based models enable confident verification of the fuel assembly inventory at the currently used detection limit for partial defects, being a 30 % discrepancy between measured and predicted intensities, while some false detection occurs with the two-parameter model. The results also indicate that the gamma-spectrum-based prediction methods are accurate enough that the 30 % discrepancy limit could potentially be lowered.

Nuclear spent fuel parameter determination using multivariate analysis of fission product gamma spectra

In this paper, we investigate the application of multivariate data analysis methods to the analysis of gamma spectroscopy measurements of spent nuclear fuel (SNF). Using a simulated irradiation and cooling of nuclear fuel over a wide range of cooling times (CT), total burnup at discharge (BU) and initial enrichments (IE) we investigate the possibilities of using a multivariate data analysis of the gamma ray emission signatures from the fuel to determine these fuel parameters. This is accomplished by training a multivariate analysis method on simulated data and then applying the method to simulated, but perturbed, data.We find that for SNF with CT less than about 20years, a single gamma spectrum from a high resolution gamma spectrometer, such as a high-purity germanium spectrometer, allows for the determination of the above mentioned fuel parameters.Further, using measured gamma spectra from real SNF from Swedish pressurized light water reactors we were able to confirm the operator declared fuel parameters. In this case, a multivariate analysis trained on simulated data and applied to real data was used.

Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

Abstract The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233U, 242Pu, 150Nd, and 133Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

ICONE23-1659 AN INVESTIGATION INTO THE FEASIBILITY OF UTILIZING THORIUM-BASED FUEL CYCLES IN A SMALL FLUORIDE-SALT-COOLED HIGH-TEMPERATURE REACTOR

The objective of this study is to investigate the potential use of thorium-based fuels, ThO2-UO2 and ThO2-PuO2 in particular, in a small (125 MWth) fluoride-salt-cooled high-temperature reactor (FHR) with lithium-beryllium fluoride salt (Li2Be4F) coolant. This study investigates thorium as an alternative fuel because it offers several potential benefits including enhanced proliferation resistance, lower waste radiotoxicity, and a higher abundance in nature compared to uranium. The deterministic lattice physics code WIMS was used to find the k∞ values and reactivity (fuel Doppler and coolant temperature) coefficients and to carry out fuel depletion. The results were verified using the Monte Carlo code SERPENT. For the analysis of homogeneous ThO2-UO2 fuels, calculations are performed over a range of thorium volume fractions and uranium enrichments (up to 20 wt.%) that can achieve the same total burnup of 5.07 MWd/kg (with acceptable reactivity coefficients) as a 100% UO2 4.69 wt.% enriched fuel reference case. The maximum thorium volume fraction that can be utilised to obtain same burnup is 0.668 at 20 wt.%. In ThO2-PuO2 analysis, reactor grade Pu with an isotopic vector taken from typical light water reactor (LWR) discharge fuel with initial 4.5 wt.% enrichment, 50 MWd/kg burnup is used. The volumetric proportions of ThO2 and PuO2 are varied and the impact on total burnup, plutonium incineration rates and reactivity coefficients investigated. To achieve the same total burnup, the thorium volume fraction is estimated to be ∼0.86.

Cross section for inelastic neutron ``acceleration'' by178Hfm2

The scattering of thermal neutrons from isomeric nuclei may include events in which the outgoing neutrons have increased kinetic energy. This process has been called inelastic neutron acceleration, or INNA, and occurs when the final nucleus, after emission of the neutron, is left in a state with lower energy than that of the isomer. The result, therefore, is an induced depletion of the isomer to the ground state. A cascade of several $\ensuremath{\gamma}$'s must accompany the neutron emission to release the high angular momentum of the initial isomeric state. INNA was previously observed in a few cases, and the measured cross sections were only in modest agreement with theoretical estimates. The most recent measurement of an INNA cross section was ${\ensuremath{\sigma}}_{\mathrm{INNA}}=258\ifmmode\pm\else\textpm\fi{}58 \mathrm{b}$ for neutron scattering by $^{177}\mathrm{Lu}$${}^{m}$. In the present work, an INNA cross section of ${\ensuremath{\sigma}}_{\mathrm{INNA}}=168 \ifmmode\pm\else\textpm\fi{} 33 \mathrm{b}$ was deduced from measurements of the total burnup of the high-spin, four-quasiparticle isomer $^{178}\mathrm{Hf}$${}^{m2}$ during irradiation by thermal neutrons. Statistical estimates for the probability of different reaction channels past neutron absorption were used in the analysis, and the deduced ${\ensuremath{\sigma}}_{\mathrm{INNA}}$ was compared to the theoretically predicted cross section.

Feasibility of 106Ru peak measurement for MOX fuel burnup analysis

Simulations were performed using ORIGEN-ARP to investigate 137Cs and 106Ru– 106Rh as suitable fission products for non-destructive analysis of irradiated MOX. The simulations confirm that both 137Cs and 106Ru will provide a linear correlation with burnup when exclusively applied to MOX fuel assemblies. Moreover, 106Ru can also be used in conjunction with cesium to form a ratio almost independent of enrichment and power history. Simulations were conducted using three different uranium enrichments and one MOX enrichment over a burnup range of 20,000–60,000 MWD/MTHM. Comparison of the three uranium enrichments indicates the 106Ru ratio is consistent in predicting burnup with a maximum standard deviation of 0.046. Two MOX cases were simulated confirming operational history independence of 106Ru in predicting total burnup. The 106Ru burnup ratio also has the benefit of enabling distinction between UO 2 and MOX fuel because of its significantly larger (∼11 times) fission yield from 239Pu. To investigate the detectability of 106Ru and other cesium peaks, data was collected using a HPGe detector at the Missouri S&T nuclear reactor (MS&TR) beam port. Gamma spectra were obtained immediately following reactor shutdown with the most promising spectrums obtained 3–5 h after shutdown. Even for relatively high enrichment (∼20%) fuel at MS&TR, cesium peaks were prominent and easily discernable from the intense Compton continuum. The 106Ru peak was weak, though still distinguishable from the background, suggesting that with an appropriately designed collimator, suitable detector and electronics it might be feasible to reliably measure 106Ru in even UO 2 fuel. For MOX and LEU LWR fuels one would expect a more intense 106Ru signature.

Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect II: Effects of variations of the fuel particle diameters

In order to predict the irradiation mechanical behaviors of plate-type dispersion nuclear fuel elements, the total burnup is divided into two stages: the initial stage and the increasing stage. At the initial stage, the thermal effects induced by the high temperature differences between the operation temperatures and the room temperature are mainly considered; and at the increasing stage, the intense mechanical interactions between the fuel particles and the matrix due to the irradiation swelling of fuel particles are focused on. The large-deformation thermo–elasto-plasticity finite element analysis is performed to evaluate the effects of particle diameters on the in-pile mechanical behaviors of fuel elements. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the fuel particle diameters; the effects of particle diameters on the maximum first principal stresses vary with burnup, and the considered case with the largest particle diameter holds the maximum values all along; (2) at the cladding near the interface between the fuel meat and the cladding, the Mises stresses and the first principal stresses undergo major changes with increasing burnup, and different variations exist for different particle diameter cases; (3) the maximum Mises stresses at the fuel particles rise with the particle diameters.

High burn-up capability possibilities for a new beryllium moderated water cooled natural uranium reactor

A recently conceived nuclear reactor design is evaluated here for theoretical burn-up characteristics which might support Global Nuclear Energy Partnership (GNEP) goals. This reactor uses natural uranium metal as fuel with beryllium moderation. The reactor also uses light water as a coolant. The reactor analysis in this work predicts the reactor to be capable of running at up to 4 GW-thermal for total burn-up values of approximately 1.4 × 10 3 GW-days. This is a very simple conceptual reactor design intended solely for very preliminary feasibility studies.

Burnup Determination of High Burnup and Dry Processed Fuels Based on Isotope Dilution Mass Spectrometric Measurements

Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors, and of a dry processed fuel fabricated from a spent PWR fuel and irradiated in the Hanaro research reactor. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U, Pu, 148Nd, 145Nd+146Nd, total of the Nd isotopes, 133Cs and 137Cs determinations by the isotope dilution mass spectrometric method (IDMS) by using quadrupole spikes (233U, 242Pu, 150Nd, and 133Cs). The methods involved two sequential anion exchange resin (AG 1X8 and 1X4) separation procedures and a Cs purification with a cation exchange resin (AG 50WX4) separation procedure. The results obtained by the Nd and Cs isotopes from the mass spectrometric measurement were compared with those by the ORIGEN code.

Use of zirconium and ruthenium isotope correlations in the analysis of spent nuclear fuels

An approach is proposed to use the isotope ratios93Zr/96Zr,91Zr/96Zr and101Ru/104Ru for the determination of plutonium contriubtion to the total burnup of a spent nuclear reactor fuel of low initial enrichment in235U (3.6%). This approach was verified by determining the total burnup of the fuel from the137Cs and148Nd monitors and subtracting the partial burnup due to235U from the value obtained.

Mass Spectrometric Determination of Burnup

A basic study was performed on the mass spectrometric determination of burnup for a UO2 specimen irradiated in the HFR. It has been found that: 1. By isotopic analysis of uranium, the contribution to burnup provided by the specimen through fissioning of 235U was determined to be 4,370 MWD/T. The loss of 235U through neutron capture was distinguished by the production of 236U. 2. The contribution to burnup of the specimen by fissioning of 239Pu was determined to be 3,610 MWD/T from the excess formation of fission products such as 93Zr, based on the known number of 235U fissions. 3. The total burnup of the UO2 specimen was taken to be 7,980 MWD/T, since no other isotopes of plutonium were identified.

Mass Spectrometric Determination of Burnup

A basic study was performed on the mass spectrometric determination of burnup for a UO2 specimen irradiated in the HFR. It has been found that: 1. By isotopic analysis of uranium, the contribution to burnup provided by the specimen through fissioning of 235U was determined to be 4,370 MWD/T. The loss of 235U through neutron capture was distinguished by the production of 236U.2. The contribution to burnup of the specimen by fissioning of 239Pu was determined to be 3,610 MWD/T from the excess formation of fission products such as 93Zr, based on the known number of 235U fissions.3. The total burnup of the UO2 specimen was taken to be 7,980 MWD/T, since no other isotopes of plutonium were identified.