Nuclide Concentrations Research Articles

Leaching resistance of simulated radioactive graphite waste forms prepared by SHS

In order to study the SHS immobilization effect of the radioactive graphite,the exothermic reaction(3C+4Al+3TiO2=2Al2O3+3TiC+ Q),in which Sr2+was used to simulate 90SrO,and Nd3+、Ce4+were used to simulate An3+,An4+,respectively,was introduced in the air and N2atmosphere by SHS equipment self-designed.The addition amount of simulated radionuclide was SrO 1%~8 %,Nd2O31%,CeO21%~3%in the air and SrO 1%~8 %,Nd2O31%,CeO21%~2%in N2atmosphere,respectively.The leaching experiments of waste forms prepared were conducted in synthetic sea water at 40 ℃ and 70 ℃.X-ray diffractometer(XRD)and inductively coupled plasma source mass spectrometer were used to test and analyze the phase composition and leaching data of these samples.The anti-leaching properties were studied on the basis of a detailed research on the phase changes of waste forms within solid solution.The results indicated that leaching concentration of Sr2+、Nd3+、Ce4+in waste forms prepared in both air and N2atmosphere increased gradually with extending immersion time,and it was higher at 70 ℃than 40 ℃.The leaching concentration of simulated nuclides was higher in waste forms prepared in N2atmosphere than in the air.

Assessment of Cs and Sr accumulation in two epiphytic species of Tillandsia (Bromeliaceae) in vitro

ABSTRACTMultiple nuclides commonly occur together and exert toxicity simultaneously, but the difference between single and combined effects of the nuclides is rarely investigated. Epiphytic Tillandsia species are efficient air pollution biomonitors, but rarely used to monitor nuclide contamination. Two Tillandsia species, that is, T. brachycaulos and T. stricta, were chosen to test their capacity to accumulate Cs and Sr. Most plants were able to endure Cs and Sr stress for a long period, which suggested these species could resist toxic elements physiologically and metabolically. With the increasing Cs or Sr concentrations, nuclide contents in both species increased significantly, indicating the potential of Tillandsia species in monitoring nuclide pollutants. However, when the plants were treated with combined nuclides, the content of each ion decreased distinctly compared to those treated with single ion, which suggested Cs and Sr influenced and inhibited each other. In addition, T. brachycaulos seemed more efficient in the uptake of Sr, while T. stricta was more efficient for Cs. Both species accumulated more Sr than Cs at low concentrations, while more Cs than Sr at high concentrations. These results indicated that the uptake of Cs and Sr was related to both the concentrations of the nuclides and the plant species exposed.

Enhanced rejection of cations by low-level cationic surfactant during ultrafiltration of low-level radioactive wastewater

Abstract The influence of cationic surfactant hexadecyl trimethyl ammonium bromide (CTAB) on the rejection of metal salts was investigated during ultrafiltration of low-level radioactive wastewater (LLRW). In the presence of CTAB below the critical micelle concentration (CMC), the rejection of nuclides significantly increased from 24%–33% to 50% for Cs(I) and above 90% for Sr(II), Co(II), and Ag(I). These increased rejections are attributed to CTAB fouling layer, which led to a more hydrophilicity and more positively-charged membrane surface. The nuclide content on the fouling layer increased with increasing CTAB concentration, and 55–80% of Co(II), Sr(II), and Ag(I) were adsorbed on the membrane at 400 mg/L CTAB. The high deposition of Ag(I) is partially due to the formation of AgBr. Based on these results, low-levels of cationic surfactant can significantly enhance the UF rejections of cations, but membrane fouling and nuclide deposition should not be ignored.

Development of a set of benchmark problems to verify numerical methods for solving burnup equations

A comprehensive set of transmutation chain benchmark problems for numerically validating methods for solving burnup equations was created. These benchmark problems were designed to challenge both traditional and modern numerical methods used to solve the complex set of ordinary differential equations used for tracking the change in nuclide concentrations over time due to nuclear phenomena. Given the development of most burnup solvers is done for the purpose of coupling with an established transport solution method, these problems provide a useful resource in testing and validating the burnup equation solver before coupling for use in a lattice or core depletion code. All the relevant parameters for each benchmark problem are described. Results are also provided in the form of reference solutions generated by the Mathematica tool, as well as additional numerical results from MATLAB.

The dependence of the nuclide composition of the fuel core loading on multiplying and breeding properties of the KLT-40S nuclear facility

The paper describes a method for determining the effective neutron multiplication factor and the breeding ratio for the KLT-40S at operating conditions. The main reactor design features are presented which are necessary for calculations. It is shown that the type of fertile nuclides has no virtual effect on the formation of the neutron flux density spectrum. The contribution of each neutron group to the fission reaction rate at the design fissile nuclide content of 18.6% is determined. The dependences of average values of macroscopic fission cross-sections, absorption of fissile nuclides and radiation capture of fertile nuclides on the fissile nuclide content in nuclear fuel are obtained. Averaging of cross-sections was carried out over neutron flux density spectra.As a result, dependences were obtained of the effective neutron multiplication factor and breeding ratio on the fissile isotope content for various fuel compositions of uranium and thorium cycles at the KLT-40S campaign start-up. In terms of the effective multiplication factor, the best result is obtained in the 232Th+233U composition when the fissile isotope content is more than 5%, and in the 238U+239Pu composition when the the fissile isotope content is less than 5%. In terms of the breeding ratio, the best result is obtained in the 232Th+235U composition when the fissile isotope content is up to 10%; if it exceeds 10%, its values are relatively the same for the 232Th+233U and 232Th+235U and 238U+235U compositions.Thus, the most efficient composition at the nuclear fuel campaign start-up is 232Th+233U with the 233U nuclide content more than 5% due to very high values of the effective neutron multiplication factor.

The CAIRN method: automated, reproducible calculation of catchment-averaged denudation rates from cosmogenic nuclide concentrations

Abstract. We report a new program for calculating catchment-averaged denudation rates from cosmogenic nuclide concentrations. The method (Catchment-Averaged denudatIon Rates from cosmogenic Nuclides: CAIRN) bundles previously reported production scaling and topographic shielding algorithms. In addition, it calculates production and shielding on a pixel-by-pixel basis. We explore the effect of sampling frequency across both azimuth (Δθ) and altitude (Δϕ) angles for topographic shielding and show that in high relief terrain a relatively high sampling frequency is required, with a good balance achieved between accuracy and computational expense at Δθ = 8° and Δϕ = 5°. CAIRN includes both internal and external uncertainty analysis, and is packaged in freely available software in order to facilitate easily reproducible denudation rate estimates. CAIRN calculates denudation rates but also automates catchment averaging of shielding and production, and thus can be used to provide reproducible input parameters for the CRONUS family of online calculators.

Influence of Low-Level Anionic Surfactant on PES Ultrafiltration Performance: Membrane Fouling and Rejection of Nuclides

Surfactants always play a special role in wastewater processes due to their amphiphilic properties. The performance of ultrafiltration was investigated for the treatment of wastewater containing low-level anionic surfactant and trace-level nuclides. Results showed that sodium dodecyl benzene sulfonate (SDBS) below the critical micelle concentration (CMC) caused significant effects on membrane fouling and rejection of nuclides. The membrane flux decreased at SDBS concentrations below the CMC but increased at the concentrations near the CMC. The phenomenon was caused by two distinct effects of SDBS, pore blocking by the monomers and enhancement of nuclide scaling caused a decrease in flux, while hydrophilic modification of the membrane surface by micelles caused an increase in flux. The nuclides alone had no significant effect on membrane fouling, but the flux decreased upon an increase in nuclide concentration when coexisting with SDBS. After the addition of low-level SDBS, the rejections of nuclides increased sharply from 20–30 to 60–98 %. The rejections of Sr(II) and Co(II) were higher than those of Ag(I) and Cs(I) due to stronger complexation of SDBS with divalent cations compared with monovalent cations. Deposition of nuclides increased with the addition of SDBS and with increasing of nuclide concentration, resulting in more radioactive solid waste production and more frequent replacement of membrane module.

First results for fluid dynamics, neutronics and fission product behavior in HTR applying the HTR code package (HCP) prototype

To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT lead to new cross sections, which are fed back into MGT-3D. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT-3D. Comparisons will be shown against data generated by SERPENT and the legacy codes VSOP99/11, NAKURE and FRESCO-II.

Edge effects on the characteristics of uranium diffusion on graphene and graphene nanoribbons**Project supported by the International S & T Cooperation Program of China (Grant No. 2014DFG60230), the National Natural Science Foundation of China (Grant Nos. 91326105, 21306220, and 21501189), and the Strategic Priority Research Program of the Chinese Academy of Sciences (Grant No. XDA02040104).

The first principles density-functional theoretical calculations of U adatom adsorption and diffusion on a planar graphene and quasi-one-dimensional graphene nanoribbons (GNRs) are performed. An energetic preference is found for U adatom diffusing to the hollow sites of both graphene and GNRs surface. A number of U distinctive diffusion paths either perpendicular or parallel to the ribbon growth direction are examined. The edge effects are evidenced by the calculated energy barriers of U adatom diffusion on armchair and zigzag nanoribbons surfaces. The calculation results indicate that the diffusion of U adatom from the inner site toward the edge site is a feasible process, particularly in zigzag GNR. It is viable to control the initial morphology of nuclear carbon material to retard the diffusion and concentration of nuclides.

Linking subaerial erosion with submarine geomorphology in the western Ionian Sea (south of the Messina Strait), Italy

AbstractSediment supplied by continental sources is commonly suspected to have exerted a strong influence on the development of canyons and other morphological features on the continental slopes, but rarely is the sediment supply known sufficiently quantitatively to test this link. Here, we outline an area where offshore morphology, in the western Ionian Sea, may be linked to estimated sediment fluxes produced by subaerial erosion in NE Sicily and SW Calabria. Shelves in this area are narrow (<1 km), and the bathymetry shows that rivers and adjacent submarine channels are almost directly connected with each other. Integrated topographic analyses were performed on a merged digital elevation model (DEM) of ASTER data for subaerial topography and multibeam sonar data for submarine bathymetry. Spatial variations in sediment fluxes from onshore erosion were assessed using a variety of methods, namely: long‐term sediment flux from Pleistocene uplift rates, decadal sediment flux from landslide occurrences and published long‐term exhumation rates from 10Be cosmogenic nuclide concentrations. Submarine channels associated with rivers delivering larger sediment fluxes have broad channels, high relief and smooth concave‐upward longitudinal profiles. Conversely, submarine channels that lie offshore small‐flux rivers have straight longitudinal profiles, low relief and steep gradients. Where river catchments supply a greater sediment flux offshore, shelves tend to be wider (ca. 400 m) and submarine channels have gentler gradients. In contrast, where catchments supply less sediment flux, shelves are narrow (250–300 m) and offshore channel gradients are steeper. The variation of submarine morphology with tectonic uplift rate was also studied, but we find that, unlike onshore terrains where tectonics is commonly an important factor influencing channel morphology, in the submarine landscapes, sediment flux appears to dominate here.

Cenozoic rejuvenation events of Massif Central topography (France): Insights from cosmogenic denudation rates and river profiles

Abstract The French Massif Central is a part of the Hercynian orogenic belt that currently exhibits anomalously high topography. The Alpine orogenesis, which deeply marked Western European topography, involved only marginally the Massif Central, where Cenozoic faulting and short-wavelength crustal deformation is limited to the Oligocene rifting. For this reason the French Massif Central is a key site to study short- and long-term topographic response in a framework of slow tectonic activity. In particular the origin of the Massif Central topography is a topical issue still debated, where the role of mantle upwelling is invoked by different authors. Here we present a landscape analysis using denudation rates derived from basin-averaged cosmogenic nuclide concentrations coupled with longitudinal river profile analysis. This analysis allows us to recognize that the topography of the French Massif Central is not fully equilibrated with the present base level and in transient state. Our data highlight the coexistence of out-of-equilibrium river profiles, incised valleys, and low cosmogenically derived denudation rates ranging between 40 mm/kyr and 80 mm/kyr. Addressing this apparent inconsistency requires investigating the parameters that may govern erosion processes under conditions of reduced active tectonics. The spatial distribution of denudation rates coupled with topography analysis enabled us to trace the signal of the long-term uplift history and to propose a chronology for the uplift evolution of the French Massif Central.

Parametric studies of the PWR fuel assembly modeling with Monte-Carlo method

In this paper we present results of studies concerning modeling of critical nuclear systems performed by continuous energy Monte Carlo burnup code. We consider a geometry model of fuel assembly for Pressurized Water Reactor neutronic simulations with depletion calculated in the infinite medium. A variety of modeling parameters and common simplifications concerning geometry complexity have been investigated in order to observe their impact on key reactor parameters like: neutron multiplication factor, nuclide concentrations, conversion ratio and distribution of generated power. Comparison of different factors helps to draw conclusions about their significance and to choose the reliable modeling approach. The presented results should be valuable for a beginning reactor researcher, but the specialists will also find a part of results interesting due to their seldom direct consideration in literature. The conclusions are general and valid not only in case of Monte Carlo simulations.We found that the lack of reactivity control in the model biases the concentration of 235U by more than 10% at end of irradiation. Next, calculation time steps longer than 50days lead to average discrepancy of neutron multiplication factor of 100pcm or more. Axial discretization of fuel requires more than 20 burnable zones to provide asymptotic behavior of burnup calculation. Lack of thermal collision tables results in underestimation of minor actinides concentrations of the order of 10% at low burnup and several percent at the end-of-life. The discrepancies caused by: the radial discretization of fuel pins, the presence of the clad–pellet gap, the applied precision of Monte Carlo calculations and the separate treatment of fuel rods, have been observed at the end-of-life to lead to a maximum relative difference in concentrations of 235U and 239Pu below 1%. Nevertheless, some studies should take also these effects into account.

Monte Carlo burnup in HTR system with various TRISO packing

Monte Carlo codes allow for the calculation of precise representation of complex HTR geometry including the double heterogeneity of a fuel. Various approaches to coated fuel particles (TRISO) placement can be applied and the impact on criticality calculations have been already assessed by several authors. Because static neutron transport simulations are affected by different fuel lattice modeling, it can be expected that the systematic discrepancies should appear during burnup calculation as well.In this paper we compare the impact of different TRISO placement and packing factor in the model on the results of the depletion simulation. For numerical tests we chose simplified model of the fuel block from High Temperature Reactor of prismatic type and the continuous energy Monte Carlo burnup code MCB5. Burnup dependent discrepancies of kinf and nuclide concentrations have been successfully observed. Five arrangement models of fuel kernels have been compared in our work: cuboidal lattice truncated by fuel rod, cubic and hexagonal lattices without rod truncation, cubic lattice with stochastic geometry option and rod filled with one layer of randomly located particles. We put emphasis on the constant mass of fuel kernels in the system and correlate obtained results with physical and statistical aspects of TRISO packing so as to explain observed discrepancies.

On the potential for regolith control of fluvial terrace formation in semi-arid escarpments

Abstract. Cut–fill terraces occur throughout the western Andes, where they have been associated with pluvial episodes on the Altiplano. The mechanism relating increased rainfall to sedimentation is, however, not well understood. Here, we apply a hillslope sediment model and reported cosmogenic nuclide concentrations in terraces to examine terrace formation in semi-arid escarpment environments. We focus on the Pisco river system in western Peru in order to determine probable hillslope processes and sediment transport conditions during phases of terrace formation. Specifically, we model steady-state and transient hillslope responses to increased precipitation rates. The measured terrace distribution and sediment agree with the transient predictions, suggesting strong climatic control on the cut–fill sequences in western Peru primarily through large variations in sediment load. Our model suggests that the ultimate control for these terraces is the availability of sediment on the hillslopes, with hillslope stripping supplying large sediment loads early in wet periods. At the Pisco river, this is manifest as an approximately 4-fold increase in erosion rates during pluvial periods. We suggest that this mechanism may also control terrace occurrence other semi-arid escarpment settings.

Quantification of subaerial and episodic subglacial erosion rates on high latitude upland plateaus: Cumberland Peninsula, Baffin Island, Arctic Canada

Long-term rates of subaerial and episodic subglacial erosion by predominately cold-based ice cover are determined for tors on weathered plateaus on Cumberland Peninsula. By measuring terrestrial cosmogenic nuclide concentrations in differentially weathered surfaces on a given tor, we reconstruct the complex exposure and erosion history involving recurring cold-based ice cover. The style and rate of subaerial and subglacial erosion at multiple tor sites on Cumberland Peninsula are assessed with a Monte Carlo approach that computes plausible exposure histories based on a proxy record of global ice volume. Constant subaerial erosion rates by weathering are low (<2 mm ka−1), corroborated by nuclide concentrations measured on two tors located on coastal ridge crests that have likely never been glaciated. Summit plateaus intermittently covered by cold-based ice throughout the Quaternary have experienced episodic subglacial erosion by plucking of fractured bedrock with rates between 1 and 16 mm ka−1. Variation of episodic erosion rates is associated with topographic position of the sampled tors and bedrock fracture density. Most of the tors were last glacially plucked in pre-ultimate glaciations, not during the Wisconsinan glaciation. Furthermore, the new approach provides evidence for the extent of ice coverage during the late Wisconsinan, which is significant if no erratics are available for exposure dating. Despite late Pleistocene intervals of ice cover and glacial plucking, tor-studded landscapes of Cumberland Peninsula are of considerable antiquity.

A comparative study of40K versus137Cs uptake as chemical analogs by vegetable plants at different concentrations of these nuclides in soil near the 30-km Chernobyl zone

137 Cs and 40 K uptake by vegetable plants was examined in a number of places in Belarus located at the outer perimeter of the 30-km Chernobyl zone. Differences in the uptake of 40 K and 137 Cs by plant species were not found to be related to the radionuclide content in the soil, but are essentially related to plant features. Vegetable plants were found to accumulate 40 K more actively than 137 Cs. For 40K and 137 Cs, the ratio of the nuclide content in plants to their content in soil shows a tendency to decrease with increasing nuclide concentration in soil.

ISOGEN: Interactive isotope generation and depletion code

ISOGEN is an interactive code for solving first order coupled linear differential equations with constant coefficients for a large number of isotopes, which are produced or depleted by the processes of radioactive decay or through neutron transmutation or fission. These coupled equations can be written in a matrix notation involving radioactive decay constants and transmutation coefficients, and the eigenvalues of thus formed matrix vary widely (several tens of orders), and hence no single method of solution is suitable for obtaining precise estimate of concentrations of isotopes. Therefore, different methods of solutions are followed, namely, matrix exponential method, Bateman series method, and Gauss–Seidel iteration method, as was followed in the ORIGEN-2 code. ISOGEN code is written in a modern computer language, VB.NET version 2013 for Windows operating system version 7, which enables one to provide many interactive features between the user and the program. The output results depend on the input neutron database employed and the time step involved in the calculations. The present program can display the information about the database files, and the user has to select one which suits the current need. The program prints the “WARNING” information if the time step is too large, which is decided based on the built-in convergence criterion. Other salient interactive features provided are (i) inspection of input data that goes into calculation, (ii) viewing of radioactive decay sequence of isotopes (daughters, precursors, photons emitted) in a graphical format, (iii) solution of parent and daughter products by direct Bateman series solution method, (iv) quick input method and context sensitive prompts for guiding the novice user, (v) view of output tables for any parameter of interest, and (vi) output file can be read to generate new information and can be viewed or printed since the program stores basic nuclide concentration unlike other batch jobs. The sample problems are chosen to serve two purposes, namely to validate the results of the code against problems where the analytical solution is possible, and the other is to demonstrate the use of particular solution method adopted for solving the problem. Besides spent fuels, results are validated for many of the useful deduced parameters of practical interest such as radioactivity, thermal power, alpha activity, neutron emission rate, and photon emission spectrum. These parameters are of utmost important in handling spent fuels, in waste disposal, in fuel management, in radiation shielding and many other areas of nuclear fuel cycle facilities.

Re-evaluation of spent nuclear fuel assay data for the Three Mile Island unit 1 reactor and application to code validation

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15×15 design with an initial enrichment of 4.013wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via high performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. The recent measurements performed by ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. These measurements serve to improve confidence in the data, to verify reported uncertainties, and to investigate previous anomalies noted in the plutonium measurements. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.

A study of the natural radioactivity and radon exhalation rate in some cements used in India and its radiological significance

The presence of natural radioactivity and radon exhalation from building materials contribute to the radiation dose received by human. So, it is essential to evaluate the activity levels of the primordial radionuclides (238U, 226Ra, 232Th and 40K) present in the building materials for the assessment of natural radiation dose. Cement is one of the major component of the building materials and is used on a large scale. In the present study, the commercially available cement samples of fifteen different brands were used to study the radon exhalation rate and activity concentration of 238U, 226Ra, 232Th and 40K nuclides. The radon exhalation rate was measured using the can technique, while the concentration of radionuclides content was determined by using gamma ray spectroscopy. The radon exhalation rate from different brands of cements was found in the range from 1.56 to 13.1 mBqkg−1h−1 with a mean value of 5.27 mBqkg−1h−1. The specific activity of uranium was found in the range 45.3–218.9 Bqkg−1 with a mean value of 111.2 Bqkg−1; 226Ra from 20.3 to 60.1 Bqkg−1 with a mean value of 35.8 Bqkg−1; 232Th from 18.8 to 60.1 Bqkg−1 with a mean value of 33.2 Bqkg−1 and 40K varied from 160.9 to 248.1 Bqkg−1 with a mean value of 199.1 Bqkg−1. The radiological parameters – radium equivalent activity, absorbed dose rate, annual effective dose, external hazard index, internal hazard index, gamma activity index and alpha index were also evaluated to assess the potential radiological hazard associated with these cement samples. Correlation coefficients for the different radionuclides have been evaluated and studied.

A multi-nuclide approach to constrain landscape evolution and past erosion rates in previously glaciated terrains

Cosmogenic nuclides are typically used to either constrain an exposure age, a burial age, or an erosion rate. Constraining the landscape history and past erosion rates in previously glaciated terrains is, however, notoriously difficult because it involves a large number of unknowns. The potential use of cosmogenic nuclides in landscapes with a complex history of exposure and erosion is therefore often quite limited. Here, we present a novel multi-nuclide approach to study the landscape evolution and past erosion rates in terrains with a complex exposure history, particularly focusing on regions that were repeatedly covered by glaciers or ice sheets during the Quaternary. The approach, based on the Markov Chain Monte Carlo (MCMC) technique, focuses on mapping the range of landscape histories that are consistent with a given set of measured cosmogenic nuclide concentrations. A fundamental assumption of the model approach is that the exposure history at the site/location can be divided into two distinct regimes: i) interglacial periods characterized by zero shielding due to overlying ice and a uniform interglacial erosion rate, and ii) glacial periods characterized by 100% shielding and a uniform glacial erosion rate. We incorporate the exposure history in the model framework by applying a threshold value to the global marine benthic δ18O record and include the threshold value as a free model parameter, hereby taking into account global changes in climate. However, any available information on the glacial-interglacial history at the sampling location, in particular the timing of the last deglaciation event, is readily incorporated in the model to constrain the inverse problem. Based on the MCMC technique, the model delineates the most likely exposure history, including the glacial and interglacial erosion rates, which, in turn, makes it possible to reconstruct an exhumation history at the site. We apply the model to two landscape scenarios based on synthetic data and two landscape scenarios based on paired 10Be/26Al data from West Greenland, which makes it possible to quantify the denudation rate at these locations. The model framework, which currently incorporates any combination of the following nuclides 10Be, 26Al, 14C, and 21Ne, is highly flexible and can be adapted to many different landscape settings. The model framework may also be used in combination with physics-based landscape evolution models to predict nuclide concentrations at different locations in the landscape. This may help validate the landscape models via comparison to measured nuclide concentrations or to devise new effective sampling strategies.