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.

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.

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 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.

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.

Low- and intermediate-level (L/ILW) radioactive waste produced in Switzerland consists of large amounts of 14C-containing irradiated steel. 14C will be released during the anoxic corrosion of the steel in the cementitious near field of an L/ILW repository. In this study, a corrosion experiment with irradiated steel was carried out to determine the speciation of 14C released during the corrosion process in conditions similar to those anticipated in the near field of a cement-based repository. The development of the experimental setup, including installation of the reactor and development of suitable analytical methods based on compound-specific 14C analysis with accelerator mass spectrometry (CSRA AMS) is reported. Time-dependent increase in the total content of 14C-bearing organic compounds in solution (TO14C) was determined by AMS and the main organic corrosion products that are 14C-bearing formate, acetate and lactate were identified by CSRA AMS after a pre-concentration step. The concentration of the 14C-bearing organic compounds was found to be very low (fmol to pmol 14C/L). Stable carbon compounds were identified and quantified while the source of stable carbon in the system has not yet been identified and the temporal evolution of the concentration of these carbon species is presently not understood.

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 study of the mechanisms of radiocarbon (14C) release from different types of radioactive waste in final disposal conditions (organic, inorganic species, liquid, or gaseous phase) is an important aspect for the definition of waste acceptance criteria in a repository. The goal of this work is to determine the behavior of activated stainless steel, under standard leaching tests performed at El Cabril laboratory, with the aim of determining the retention capacity of the activated stainless steel, and therefore to assess the treatment and the potential additional barriers necessary for the fulfillment of waste acceptance criteria relevant to the El Cabril surface repository, Spain.

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.

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.

The gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is 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. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

The total 14C content and its speciation (inorganic/organic) were measured in spent ion exchange resins (SIERs) received from Cernavoda Nuclear Power Plant (NPP). Also, 14C release from SIERs was investigated by desorption tests carried out in alkaline solution relevant for cementitious environment disposal. The method used for total 14C measurement consists of combustion in oxygen-rich atmosphere, while for speciation measurements, both in SIERs and in desorption solutions, an analytical method based on acid stripping and wet oxidation was applied. Around 97% from the total 14C inventory measured on the Cernavoda SIERs (33.7 kBq/g) was found to be in inorganic form and only 7% as organic 14C. Under alkaline conditions, 14C could be released both as gaseous and as soluble species: from the total 14C present in the SIERs samples around 7% was released as inorganic 14C in the gas phase and 79% as dissolved species (mainly as inorganic 14C). These percentages were obtained for unconditioned SIERs in NaOH solution. The SIERs will be immobilized in a suitable matrix for disposal, and the presence of Ca ions dissolved in cement pore water favor precipitation of 14C and consequently the amount of 14C released from disposal area should be lower.

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.

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).

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.

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.

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).

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.

The understanding of the 14C behavior in waste packages could lead, in the Spanish context, to a revision of the management strategies for radioactive waste and a revaluation of the near surface repository devoted to the disposal of waste containing this radionuclide in high concentrations. To achieve this objective, and in the context of the EU project Carbon-14 Source Term (CAST), the authors of the work presented in this paper have performed leaching experiments with irradiated graphite considering two different scenarios. One, in which the leaching solution simulates some of the expected conditions in a repository where a granite/bentonite mixture has been used as backfill material, and the other, using deionized water as a high efficiency chemical removal agent and for comparison purposes. The analytical approach to measure the release rate and speciation of 14C from irradiated graphite samples in the aqueous and gaseous phase is also described. The main results obtained shows that, after 359 days of leaching, no 14C activity was detected above the detection limits, and only leaching rates regarding beta-gamma emitters were observed: 2×10–6 cm/day for 137Cs and 1×10–5 cm/day for 60Co in granite/bentonite water media and 4×10–6 cm/day for 137Cs in pure water.