Luristan Iron Mask Swords have been recovered mostly from illegal diggings in the 1920s. The about 90 known objects are characterized by a disk-shaped pommel on the top of the handle with two mounted bearded heads on two sides. According to the similarity in form and radiocarbon (14C) measurements on two swords from museum collections, an overall short production period was assumed around 1000 BC (Moorey 1991; Rehder 1991). Here we present the results of metallurgical analysis and 14C measurements for three newly acquired Luristan swords, which were donated to the Royal Museums of Art & History, Brussels. Metallurgical analysis indicates an iron production via the bloomary furnace technique. Analyzed samples show large slag inclusions (Fayalite, Wüstite, glass) within a mostly ferritic and pearlitic iron. The carbon contents varied between 0.2 wt% to around 0.8 wt%). 14C measurements on thermally extracted carbon give 14C ages between 2800 BP–3360 BP (calibrated ∼1745 BC–900 BC). The reliability of the 14C measurements are discussed with respect to external (contamination during handling) and intrinsic contamination (e.g. fossil carbon sources during manufacture).

Radiocarbon (14C) is one of the key radionuclides for the performance and safety assessment of a radioactive waste disposal, due to its high activity concentration in waste materials from the nuclear cycle and to its mobility. The measurement of the 14C content in spent ion exchange resins from nuclear reactors is important for the safety assessment of the disposal concept and for the choice of the appropriate treatment/disposal method. Ion exchange resins are commonly used in nuclear reactors as filters for the purification of process liquids or wastes streams and they retain molecules containing radioactive isotopes, among which is 14C. Their efficiency, both as filters and as waste containers, is strictly connected with the morphology. The preservation of spherical shape upon aging is one of the key parameters for their quality assessment and for the evaluation of the potential release of 14C during storage conditions. In this study, the change in IERs morphology during storage periods has been investigated in order to verify correlation with 14C release. Both brand new and aged specimens have been studied in order to assess the quality of the resins after 10 yr of storage and to contribute to the understanding of 14C release mechanisms.

The European Commission CAST project (CArbon-14 Source Term) aimed to develop understanding of the potential release mechanisms of carbon-14 (radiocarbon, 14C) from radioactive waste materials under conditions relevant to waste packaging and disposal to underground geological disposal facilities. The project focused on the release of carbon-14 as dissolved and gaseous species from irradiated metals (steels, Zircaloys), from spent ion-exchange materials and from irradiated graphites. This paper provides an overview of the CAST project and its output. It also acts as an introduction and scene-setter to the other papers in this special edition of Radiocarbon.

The radiocarbon (14C) content of irradiated graphite is the most important problem for the management of Spanish irradiated graphite (Vandellós I NPP) as L&ILW, due to this material exceeding the maximum 14C inventory for the C.A. El Cabril repository. Therefore, the encapsulation of graphite in an impermeable matrix and making an appropriate waste form are indicated as potential management options to be studied. The conversion of the graphite to a long-term stable glass matrix, called IGM (impermeable graphite matrix), uses a long-term stable inorganic binder which additionally encloses the graphite pore system. The world’s first IGM samples made with irradiated graphite have been manufactured in CIEMAT facilities. The durability of the matrix is investigated in leaching experiments in deionized water and granitic bentonite water. The results show that ∼0.05% of 14C is leached. A species of organic carbon was found as formate and oxalate (∼10–1 mg/L). CO was detected as volatile specie in both media in the first leaching steps; for deionized water (∼3.101 mg/L) and in granitic bentonite water (ranging 1.101–3.101 mg/L). These low values demonstrated the durability of the IGM glass matrix for final disposal.

Carbon-14 (radiocarbon, 14C) is an important radionuclide in the inventory of radioactive waste in many disposal programs due to its significant dose contributions in safety assessments for geological repositories. Activated steels from nuclear reactors are one of the major sources of 14C. Knowledge of 14C release from steel wastes and its chemical form (speciation) is limited giving rise to uncertainty regarding the fate of 14C and a conservative treatment in assessment calculations. In this work, we summarize and make a synthesis of selected results from Work Package 2 of the EU CAST project aiming to improve understanding of 14C release related to steel corrosion under repository-relevant conditions. The outcome of the experiments is discussed in the context of the long-term evolution of a repository and its potential consequences for safety assessment.

The potential 14C (carbon-14, radiocarbon) flux from disposal of 14C containing waste into air is compared with the natural 14C emanation rate from soil in order to put the 14C hazard potential from disposal of this waste in perspective with the 14C exposure from cosmogenic origin. Chemical corrosion of neutron irradiated metals, steel and Zircaloy, is bounded by diffusion of water through a thermodynamically stable metal-oxide layer and dissolution of this metal-oxide in a nuclear plant. Many countries process radioactive waste for disposal using cementitious materials, an acknowledged end-point management technique for this waste. The metal-oxides are also stable when these waste forms are embedded in cementitious materials. The 14C release rate from this Zircaloy at these alkaline and reducing conditions is comparable to the natural 14C emanation rate from soil into air. Neutron irradiated graphite and spent ion exchange resins are chemically inert and therefore other release mechanisms need to be assumed. Radiolytic corrosion is used to determine the 14C release rate from this graphite. Moreover, ion exchange—with ingressing anionic species that have a higher affinity than contained anionic 14C—is proposed as a release mechanism for these resins.

There are two units with RBMK-1500 type reactors at the Ignalina Nuclear Power Plant (Ignalina NPP) in Lithuania where graphite was used as a neutron moderator and reflector. These reactors are now being decommissioned, and Lithuania has to find a solution for safe irradiated graphite disposal. It cannot be disposed of in a near surface repository due to large amounts of 14C (radiocarbon, carbon-14); thus, a deep geological repository (DGR) is analyzed as an option. This study had the aim to evaluate 14C migration from the RBMK-1500 irradiated graphite disposed of in a potential DGR in crystalline rocks taking into account the outcomes of the research performed under the collaborative European project CAST (CArbon-14 Source Term) and to identify the potential to reduce the conservatism in the assumptions that was introduced in the lack of data and led in the overestimated 14C migration. The information gathered during the CAST project was used to model 14C transport in the near field by the water pathway and to perform uncertainty analysis. The study demonstrated that more realistic assumptions could reduce the estimated 14C flux from the near field by approximately one order of magnitude in comparison with the previous estimations based on very conservative assumptions.

Carbon-14 (radiocarbon, 14C) is a long-lived radionuclide (5730 yr) of interest regarding the safety for the management of intermediate level wastes (ILW). The present study gives an overview of the release of 14C from irradiated Zircaloy cladding in alkaline media. 14C is found either in the alloy part of Zircaloy cladding due to the neutron activation of 14N impurities by 14N(n,p)14C reaction, or in the oxide layer (ZrO2) formed at the metal surface by the neutron activation of 17O from UO2 or (U-Pu)O2 fuel and water from the primary circuit in the reactor by 17O(n,α)14C reaction. Various irradiated and unirradiated Zircaloys have been studied. The total 14C inventory has been determined both experimentally and by calculations. The results seem to be in good agreement. Leaching experiments were conducted in alkaline media for several time durations. 14C was mainly released as carboxylic acids. Further, corrosion measurements were performed by using both hydrogen measurements and electrochemical measurements. The corrosion rate (CR) ranges from a few nm/yr to 100 nm/yr depending on the surface conditions and the method used for measurement. From a safety assessment point of view, the instant release fraction (IRF) was determined on irradiated Zircaloy-2. The results showed that the 14C inventory in the oxide was significantly below the 20% commonly used in safety case assessments.

A preliminary evaluation of gaseous radiocarbon (14C) behavior under geological repository conditions for Italian radioactive high level waste-long-lived and intermediate level waste disposal has been performed. Although in Italy there is still no defined project for a geological disposal facility, current work may support future safety assessment studies for a hypothetical future repository in deep salt rock. In the Italian context of radioactive waste, the percentage of 14C bearing waste to be disposed in a possible geological repository is low; irradiated graphite is the most important radiological source. Data about the radiological inventory has been collected to simulate production and migration of gaseous 14C in a hypothetical geological repository. Three different conceptual models have been developed and simulated. The first model has considered a preliminary evaluation of the radiological impact referred to the whole inventory; the second and third model have evaluated the impact only due to the irradiated graphite. A preliminary sensitivity analysis has been carried out, highlighting the importance of geometry and of distribution coefficients (Kd) in materials used to seal the disposal underground facility. Results show the possibility to correlate the Kd values, the volume and the location of the sealing materials to the amount of 14C migrating toward the surface.

The chemical form of 14C released from irradiated stainless steel is a key parameter in the safety assessment of the subsurface disposal system in Japan. In this study, to identify the chemical form of the released carbon, unirradiated stainless-steel powders, which were found to be water-atomized powders with a silicon oxide film, were immersed in NaOH and HCl solutions under low-oxygen conditions for approximately 25 days. The results showed that the main chemical forms of the carbon were colloidal carbon in the NaOH solution and colloidal carbon and formic and acetic acids in the HCl solution. Almost no hydrocarbons were detected in both solution systems. Concerning the source of the colloidal carbon and carboxylic acids, the hypothesis that carbon in the oxide layer is released is considered to be reasonable. The very small amounts of hydrocarbons generated prevented us from discussing the source of the hydrocarbons. To validate the hypothesis and obtain further information on the hydrocarbons, additional experiments are necessary. In particular, for long-term safety assessment, it is important to determine whether the colloidal carbon, carboxylic acids and hydrocarbons are continuously released during the corrosion process. Therefore, information on the temporal evolution of the carbon should be obtained.

Radiocarbon (14C or carbon-14, half-life 5730 yr) is a key radionuclide in the assessment of the safety of a geological disposal facility (GDF) for radioactive waste. In particular, the radiological impact of gaseous carbon-14 bearing species has been recognized as a potential issue. Irradiated steels are one of the main sources of carbon-14 in the United Kingdom’s radioactive waste inventory. However, there is considerable uncertainty about the chemical form(s) in which the carbon-14 will be released. The objective of the work was to measure the rate and speciation of carbon-14 release from irradiated 316L(N) stainless steel on leaching under high-pH anoxic conditions, representative of a cement-based near field for low-heat generating wastes. Periodic measurements of carbon-14 releases to both the gas phase and to solution were made in duplicate experiments over a period of up to 417 days. An initial fast release of carbon-14 from the surface of the steel is observed during the first week of leaching, followed by a drop in the rate of release at longer times. Carbon-14 is released primarily to the solution phase with differing fractions released to the gas phase in the two experiments: about 1% of the total release in one and 6% in the other. The predominant dissolved carbon-14 releases are in inorganic form (as 14C-carbonate) but also include organic species. The predominant gas-phase species are hydrocarbons with a smaller fraction of 14CO (which may include some volatile oxygen-containing carbon-species). The experiments are continuing, with final sampling and termination planned after leaching for a total of two years.

The total 14C content and its partition between inorganic and organic species were measured on irradiated Zy-4 samples from a CANDU spent fuel rod transferred from Cernavoda Nuclear Power Plant (NPP). Long-term leaching tests and accelerated corrosion tests were carried out to measure the 14C release and corrosion rate, respectively, in chemical conditions relevant to cementitious environment. Experimentally measured 14C inventory was compared to the theoretically one predicted based on the irradiation history and impurity content of Zy-4 by means of ORIGEN computations. CANDU SF claddings have a 14C content of around 2 × 104 Bq/g of Zy-4, mainly as organic compounds (more than 99%). The total 14C content measured by acid dissolution/wet oxidation method is in good agreement with the value estimated by ORIGEN simulations for an average burn-up of 7 MWd/kgU. The total 14C released as dissolved species after 18 days and 18 months of Zy-4 immersing in alkaline solution are similar, indicating that a small amount of 14C was available as instant release fraction (0.05% from the initial 14C content) followed by a very low release rate that could not be measured by liquid scintillation counting. In alkaline conditions, 14C is released predominantly (∼70%) as soluble species, but also inorganic 14C was measured as gaseous species. From the soluble 14C released during leaching test, more than 60% was found to be as organic species. Generally, corrosion rates values ranging between 46 and 130 nm/yr were measured by the linear polarization resistance method. In addition, defects and cracks were observed on the oxide layer by scanning electron microscopy (SEM) investigation.

Leaching tests were carried out in aerobic and anaerobic conditions to assess the 14C released from TRIGA irradiated graphite. Both total 14C and inorganic and organic fractions in the leachant solutions were measured. The experimental results obtained from the leaching tests confirm the low 14C release rate in alkaline environment. Less than 2% of the total 14C inventory in the specimens subject to the leaching tests was released as dissolved species. Both inorganic and organic 14C species are released in alkaline conditions, with more inorganic 14C release under aerobic conditions (around 68% of the total 14C released was released as inorganic dissolved species), and more organic 14C species in anaerobic conditions (around 65% of the total 14C released was released as organic dissolved species). Both for anaerobic and aerobic conditions, the leaching rates are high in the first days of immersion and decrease after that, indicating a two stage process: an initial quick release (less than 9 × 10–02 % of inventory/day for the first 48 days) followed by a slower release rate (around 4 × 10–03 % of inventory/day).

Samples of graphite from a RBMK-1500 reactor at the Ignalina Nuclear Power Plant from different construction elements (stack, sleeve, and bushing) were analyzed by the instrumental neutron activation analysis (INAA) method (LVR-15 experimental reactor of the Research Centre Řež, Ltd.) using the prompt gamma activation analysis (PGAA) method (Heinz Maier-Leibnitz Zentrum) and with an inductively coupled plasma mass spectrometer (ICP-MS) (CPST, Lithuania). These measurements were performed with the aim of obtaining the missing information on the impurity distribution in the RBMK-type nuclear graphite constructions as well as for intercomparison purposes, with the results measured in the graphite sleeve samples previously obtained by INAA & GDMS (Glow Discharge Mass Spectrometry) at CEA Saclay, France, and ICP-MS (CPST, Lithuania) methods. Validation of the ICP-MS method for the nuclear graphite impurity concentration determination was proven. The experimentally obtained RBMK-1500 graphite impurity values in different graphite constructions were compared with other measurements and new limits of the possible maximal concentrations of nuclear RBMK graphite impurity concentrations were obtained.

Generally, 14C is produced in all types of reactors, mainly through neutron-induced reactions with isotopes of carbon, nitrogen, and oxygen. These isotopes are present in the fuel, fuel cladding, coolant, moderator, and structural materials of the reactor. Irradiated graphite contains a range of activation products but, from the point of view of geological disposal, another radionuclide which has to be taken into account is tritium (3H or T), produced through the neutron induced reaction Li(n,α)T. In this study we have investigated the release of 14C and 3H into the solution phase from two intact samples of irradiated graphite which have been cut from the thermal column disc located near the VVR-S reactor core. To evaluate 14C and 3H long-term release in solution the graphite samples were submerged in a solution of 0.1 M of sodium hydroxide. The experimental results regarding the release of 14C and 3H from VVR-S irradiated graphite to liquid phase show that a major fraction of the total release occurs in the first months and a slower release on a long time scale. However, these results should be applied cautiously for long time prediction.

Radiocarbon (14C) is a key radionuclide in the assessment of the safety of underground geological disposal facilities for radioactive wastes, and the understanding of the 14C behavior in stainless steel may lead to a re-evaluation of the near-surface repository for the disposal of wastes containing this radionuclide in high concentrations. To achieve this objective, leaching experiments were planned for two different scenarios. The first is where the leaching solution, NaOH solution of pH ca. 12 in aerobic conditions, simulates the expected conditions in a cement-based near-surface repository over long time periods. The other one uses an acid solution of 1M H3PO4, which has been proved as a high efficiency chemical removal agent of 14C in graphite. The development of both analytical methods and protocols to measure the release of 14C from the activated steel samples and the speciation in the aqueous and gaseous phase has been undertaken as part of the EC CAST (CArbon-14 Source Term) project. Analytical methods, suitable for identifying and quantifying low molecular weight organic molecules, comprise ion chromatography (IC) and gas chromatography coupled to mass spectrometry (GC-MS); they are described for aqueous and gaseous samples, respectively. In this paper the preparation of leaching experiments to measure the release of 14C and the results obtained are described.

The Neolithic transition is a particularly favorable field of research for the study of the emergence and evolution of cultures and cultural phenomena. In this framework, high-precision chronologies are essential for decrypting the rhythms of emergence of new techno-economic traits. As part of a project exploring the conditions underlying the emergence and dynamics of the development of the first agro-pastoral societies in the Western Mediterranean, this paper proposes a new chronological modeling. Based on 45 new radiocarbon (14C) dates and on a Bayesian statistical framework, this work examines the rhythms and dispersal paths of the Neolithic economy both on coastal and continental areas. These new data highlight a complex and far less unidirectional dissemination process than that envisaged so far.

Ignalina NPP contains two Units with RBMK-1500 reactors. After shutdown, several Unit 1 systems and equipment were radiologically characterized and dismantled. The highest volume of reactor structures is attributed to the graphite stack of the reactor core, radiological characterization of which has not yet been performed. The stack can be visualized as a vertical cylinder 8 m high and 14 m diameter, made up of 2488 columns where each column is made up from several graphite blocks. The total mass of the graphite stack blocks is about 1700 tonnes. Therefore, the main goal of work reported in this paper was to estimate the inventory of 14C and other key radionuclides in the irradiated graphite by a combination of activity measurements and full 3D reactor graphite stack neutron activation modeling. Obtained results show that, based on the combination of modeling and measurement techniques, the total inventory of 14C in graphite stack is estimated at 3.22×1014 Bq at 9 years after Unit 1 reactor final shutdown. 14C activity is the highest among the analyzed radionuclides; the second highest is 60Co (~6 times lower).

The release and the speciation of carbon species from irradiated JRQ carbon steel samples, representative of the reactor pressure vessel of Belgian nuclear power plants, were studied in a saturated portlandite aqueous solution, relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. In addition, the corrosion rate (which determines the 14C release) was estimated by measuring the release of 60Co. Gas chromatography showed that during the static corrosion test, the carbonaceous species methane, carbon dioxide, ethene, and ethane were produced. Under the hypothesis that all the carbon released from the JRQ steel was transformed into carbon-base gaseous compounds, this corresponds to a corrosion rate of approximately 100 nm/yr, which is in good agreement with literature data.