Community Research and Development Information Service - CORDIS

Final Report Summary - CAST (CAST (CArbon-14 Source Term))

Executive Summary:
This overview intends to summarize the results and achievements of CAST and their implications for building confidence in the safety of geological disposal. The half-life of carbon-14 is 5730 years. In many disposal concepts for High Level Waste (HLW), there is an engineered containment period up to several hundred thousand years. The activity concentration of carbon-14 in waste is significantly reduced in this containment period. Investigation of the CArbon-14 Source Term is therefore of more interest for the Low and Intermediate Level Waste (LILW) than HLW for many countries. Experiments to measure the release of carbon-14 and carbon-14 speciation have therefore been mainly performed at room temperature. The CAST project focused on the release of carbon-14 as dissolved and gaseous species from waste. Carbon-14 can be released as an organic species i.e. carboxylic acids, alcohols, aldehydes, alkenes and alkanes or as an inorganic species i.e. carbonate, bicarbonate and carbon dioxide. All species need to be measured, in order to determine the main released carbon-14 species. The measurement of released carbon-14 species for the neutron irradiated materials is challenging due to its low chemical amount in waste and low corrosion rates in cementitious materials. Typical carbon-14 activity concentrations in neutron irradiated steel from the core of a nuclear plant are in the order of 1E5 Bq per gram solid matter and for neutron irradiated Zircaloy 1E4 Bq per gram solid matter i.e. a few ppm in neutron irradiated steel or tenths of ppm in neutron irradiated Zircaloy. The speciation of released carbon-14 species could not be quantified within CAST. These materials contain non-radioactive carbon: for example in steel, the non-radioactive carbon content is three orders in larger magnitude than the carbon 14 content. In CAST, significant progress has been made to quantify the speciation of released non-radioactive carbon from steel. The major carbon-14 species were dissociated anionic compounds from carboxylic acids in the case when these dissociated anionic compounds and gaseous organic carbon were measured; gaseous organic carbon was negligible compared to dissolved organic compounds. An analytical strategy for the quantification of released alcohols and aldehydes has been developed. For neutron irradiated steel, carbon is released as dissolved non-gaseous non-ionic compounds i.e. alcohols and aldehydes. These compounds cannot be retarded in the cementitious materials by sorption and ion exchange due to lack of electrostatic interactions.
The presence of hydrogen can influence the fate of released carbon species; the carbon-14 released as non-ionic compounds such as alcohols and aldehydes can be further reduced with hydrogen to gaseous organic carbon. Significant confidence has been raised that more than 90% of the hydrogen generated during anaerobic corrosion is picked-up during corrosion of Zircaloy at disposal conditions. No hydrogen is formed during degradation of graphite and ion exchange resins under disposal conditions, but hydrogen can be formed by radiolysis of concrete pore water by radiation from neutron irradiated graphite and spent ion exchange resins. These radiolytic hydrogen generation rates are usually neglected. Consequently, it is only for neutron irradiated steel that the potential reduction to gaseous organic carbon needs to be taken into account. Carboxylic acids are expected only to occur as dissociated anionic compounds at disposal; their presence hinders the reduction to gaseous alkenes and alkanes such as ethene, methane and ethane with hydrogen. Only in acidic media for example in acid digestion tests, this potential reduction of carboxylic acids is expected to be possible.
The OH radicals formed by radiolysis of concrete pore water can also have an influence on the fate of released organic carbon. These radicals can oxidise alcohols and aldehydes into inorganic carbon. This process may have occurred during the experimental investigations in CAST in which neutron irradiated steel and neutron irradiated Zircaloy were exposed to portlandite solutions. Gases released from these solutions were measured to be organic gaseous carbon as well as carbon dioxide. Radiolysis may be the reason why both reduced and oxidised hydrocarbons can exist simultaneously in solutions in contact with corroding metals. If the measured carbon dioxide is presumed not to be a contaminant, the carbon dioxide may have been formed by oxidation of dissolved organic compounds with OH radicals. Also, in alkaline solutions with spent ion exchange resins, carbon-14 bearing carbon dioxide has been measured. This induced oxidation of dissolved organic compounds is positive for a potential retardation in cementitious materials, since any carbon dioxide released from waste in cementitious materials will form an unstable carbonic acid that is dissociated into bicarbonate and carbonate. The high calcium content in concrete pore water and low solubility will cause precipitation. Another retention mechanism of carbonate is sorption on positively charged cementitious minerals.
Dissolved organic carbon compounds have been measured to be the major carbon-14 species released from neutron irradiated steel, neutron irradiated Zircaloy and neutron irradiated graphite. Distinction between carboxylic acids, alcohols and aldehydes is to be made. The release of non-ionic carbon compounds such as alcohols and aldehydes is therefore conservatively assumed to be the main carbon-14 species. The extent to which these species can be assumed to be retarded in engineered and natural barriers is uncertain, and in the case of natural barriers depends on the host rock properties. These neutron irradiated materials are also known to have very small mass loss rates, so the potential carbon-14 fluxes to air from these three waste types as a consequence of their deep geological disposal is expected to be negligible compared to the flux of natural carbon-14 emanating from soil.
The main origin of carbon-14 in neutron irradiated steel and Zircaloy is nitrogen. Measured and calculated carbon-14 contents are of the same order in magnitude. For neutron irradiated graphite, it depends on the chemical and temperature conditions whether nitrogen is the main contributor to carbon-14 in waste. Nitrogen-activated carbon-14 forms a loosely bound carbon-14 gaseous species that can e.g. be released during crushing of neutron irradiated graphite.
Only measured carbon-14 contents in resins that have been prepared for conditioning are representative of behaviour likely in relation to a disposal scenario, due to the high impact of the storage and drying conditions on the contained carbon-14 content. Knowledge of the contained inorganic and organic carbon content is necessary for safety assessments when organic carbon compounds are assumed to be non-retarded species. Conservatively, the speciation of carbon-14 contained by resins is assumed to be released instantaneously. Spent ion exchange resins can only contain anionic carbon-14 compounds. This containment limits the presence of organic carbon in ion exchange resins to dissociated anionic compounds of carboxylic acids such as oxalate and acetate; the only possible contained inorganic compounds are bicarbonate and carbonate. Attempts to measure release of carbon-14 in alkaline media have been made but carbon-14 has not been measured to be released in cementitious pore water. Inorganic carbon is assumed to be retarded in cementitious materials due to the precipitation with calcite i.e. fixation by high calcium amount in the concrete pore water chemistry. This retention mechanism also takes place for oxalate. Sorption of positively-charged cementitious minerals is another retention mechanism and in CAST the literature for sorption of acetate has been found. The last evaluated retention mechanism is ion exchange; formate is exchanged with sulphate in ettringite. Consequently, there is no reason in cementitious materials only to assume retention for inorganic carbon and not for organic carbon during disposal of spent ion exchange resins when processed with cementitious materials. The amount of carbon-14 that is released and not decayed within the waste package may be negligible.

Project Context and Objectives:
Carbon-14 (14C, half-life 5,730 years) is a radionuclide of importance to safety assessments. The CAST project (CArbon-14 Source Term) develops understanding of the release of 14C from radioactive wastes under conditions relevant to geological disposal.

The main objectives of CAST are:
a) To gain scientific understanding of the rate of release of 14C from the corrosion of irradiated steels and Zircaloys and from the leaching of ion-exchange resins and irradiated graphites under geological disposal conditions, its speciation and how these relate to 14C inventory and aqueous conditions;
b) Evaluate this understanding in the context of national safety assessments; and
c) Disseminate this understanding and its relevance to safety assessments to a range of interested stakeholders and provide an opportunity for training of early career researchers.

These objectives are met through seven Work Packages:
• WP1 - Coordination (led by RWM, UK)
• WP2 - Steels (led by Nagra, Switzerland)
• WP3 - Zircaloy (led by Andra, France)
• WP4 - Ion-Exchange Resins (led by CEA, France)
• WP5 - Graphite (led by RWM, UK)
• WP6 - Relevance to Safety Cases (led by Ondraf/Niras, Belgium)
• WP7 - Dissemination (led by Covra, Netherlands)

WP1 - Coordination
The main objectives for WP1 Management are to:
• Manage and coordinate the project;
• Advise and support the work package leaders;
• Assist WP7 with the CAST website;
• Collect and compile project management information;
• Distribute EC financial contributions;
• Provide quality assurance and control of project results.

WP2 - Steels
The main objectives for WP2 Steels are to:
• Develop analytical techniques to identify and quantify 14C species formed during corrosion of irradiated steels;
• Validate activation models by measuring 14C inventories in irradiated steel;
• Carry out experiments and modelling to further understand speciation and rate of 14C release from corrosion of irradiated steels under conditions relevant to deep geological repositories.

WP3 - Zircaloy
The objective of WP3 is to obtain a better understanding of 14C behaviour in waste Zr fuel claddings under disposal conditions by:
• Assessing 14C inventories in zirconium alloy metals and oxides;
• Characterising 14C release from Zr corrosion and Zr oxide dissolution;
• Determining 14C speciation under simulated disposal conditions.

WP4 - Ion Exchange Resins
The aim of WP4 is to obtain a better understanding of the 14C source term from spent ion-exchange resins (SIERs) of different origins (BWR or PWR) and release and chemical species under disposal conditions by:
• Reviewing the current status of understanding;
• Characterising the 14C inventory and its speciation;
• Undertaking experiments to measure 14C release to gas and solution.

WP5 - Graphite
The objective of this work package is to understand the factors determining release of 14C from irradiated graphite (i-graphite) under geological disposal conditions by:
• Determining the 14C inventory and concentration distribution in i-graphites and factors that may control these;
• Measuring the rate and speciation of 14C release to solution and gas from i-graphites in contact with aqueous solutions;
• Determining the impact of selected waste treatment options on 14C release and relating this to the nature of 14C in i-graphite.

WP6 - Relevance to Safety Cases WP6 integrates results with the aim of:
• Combining the results of WP 2 to 5 to deliver sound scientific basis and safety relevant information;
• Considering the results in the context of safety cases;
• Identifying commonalities and differences between national programmes;
• Providing conclusions and recommendations over possible future studies.

WP7 - Dissemination
The objective of WP7 is to disseminate information about CAST and its results to target groups. This is provided via a public project website and newsletters and scientific publications. Two training courses will be held to train early career researchers and two workshops to engage key stakeholders.
Project Results:
WP1 - Coordination
Four General Assembly Meetings have been held: London, 2013; Brussels, 2014; Bucharest, 2015; and Lucerne (October 2016). A CAST symposium was held in Lyon, France, in January 2018. This was open to CAST and non-CAST participants.
The CAST coordinator has approved all CAST reports for publication, and has overseen the overall management of the project.

The results from CAST have been captured in overall synthesis reports (Work Packages 2-7), summarising the results and achievements of the whole project. WPs 2 to 5 each produce a final work package synthesis report for public dissemination describing: a) the key results from the work package; b) the relevance of the scientific issues addressed for the repository concepts of the countries involved; c) how the results improve the understanding of the evolution of the repository concepts and wastes; and d) the implications for data input in PA calculations and for safety case arguments for the concepts considered (with WP6). WP6 will deliver a final report for public dissemination.

Work Package 2: Final synthesis report on results from Irradiated Steels studies (See Deliverable D2.18 for further information and references)
In Work Package (WP) 2 of the CAST project twelve partners from different countries came together to study the release of 14C from activated steels, e.g. from reactor internals or the reactor pressure vessel. The objectives of WP2 were to develop suitable analytical techniques for measurement of 14C speciation, to validate existing activation models, to measure the 14C release and speciation in leaching experiments and finally to advance our understanding of 14C behaviour released upon steel corrosion.
In task 1 of WP2 a literature research was made, and a State-of-the-Art report was published (D2.1). The outcome was that the corrosion process of steels is well understood. The corrosion rates under a variety of conditions are known and their uncertainties are low. A good understanding exists on the prevalence of carbon in steels forming different kinds of carbides. In only a limited number of experiments the release of stable carbon from steels was studied and even less studies exist on the release of 14C from activated steels. Dissolved and gaseous compounds were found in both cases.
In task 2 analytical methods were developed to measure the speciation of 14C in gaseous and liquid samples at low concentrations due to the expected slow release. Methods comprise ion chromatography for dissolved species and gas chromatography coupled with mass spectrometry for gaseous species. For the low-level detection, accelerated mass spectrometry was applied. Other techniques involved selective oxidation with subsequent capture. A method for separation of other radionuclides from 14C was developed.
Task 3 involved a series of leaching experiments using samples of activated stainless steels, mild steel and some other, non-activated materials. Most experiments applied alkaline, anaerobic conditions but some tests involved neutral and acidic pH values as well as aerobic conditions. The 14C speciation could not be measured in all experiments. In many cases, a fast initial release of both 12C and 14C to the gas and liquid phase could be observed after immersion of the samples into the leaching solution. Later, within some days to a few weeks, the release rate decreased significantly. As a first approximation, this release rate seems to correspond to the corrosion rate of steel, which is, however, subject to rapid changes in the early stage of the experiment. The conclusion of a 14C release congruent with the steel corrosion is therefore subject to high uncertainty. The measured speciation of 14C comprises both organic and inorganic compounds in the liquid phase. Methane and minor contributions of CO were found in the gas phase. No systematic differences in the behaviour of 12C and 14C could be observed. Since no rigorous comparison of the behaviour of both isotopes could be made, it cannot be excluded that differences exist. The 14C concentration in selected samples has been estimated using different activation models and compared to measured activity concentrations. An uncertainty factor of 3 to 4 has been found whereby the measured concentrations were higher than estimated.
In task 4 a critical discussion of the results and a synthesis was made. When bounding the possible long-term behaviour, the system evolution for two different situations is considered: an early and a late resaturation, corresponding to a crystalline host rock with higher water influx and a clay host rock with extremely low water supply, respectively. Corrosion rates remain low for both situations. In the early stage, the release of 14C as carbonate or carboxylic acids can be expected according to the experiments. As the resaturation process, also in the case of high water supply, will take much longer in reality than simulated in the experiments, the observed fast initial release will stretch over a longer period of time.
The long-term release rate and speciation of 14C was experimentally not accessible within the time frame of the CAST project. There are, however indications, what the long-term evolution might be. It could be observed that samples featuring a higher dose rate (Wood/NRG) tend to produce more carbonate in the liquid phase and CO in the gas phase. In an experiment using a piece with a lower dose rate (PSI), the solution was dominated by carboxylic acids and the gas phase by methane. The conclusion is drawn that the higher dose rate and the presence of oxidants due to radiolysis might influence the speciation of carbon after release from activated steel. A later reduction of the dose rate in a time frame of several 100 to 1,000 years due to the decay of some short-lived radionuclides, e.g. 60Co, would mean that the effect of radiolytically produced radicals would cease and the speciation would be shifted to compounds featuring no oxygen-containing functional groups. In such a case, hydrocarbons would be mainly expected. That would mean that a higher percentage of produced 14C compounds would be expected in the gas phase. This suggestion needs further experimental studies to be supported.

Work Package 3: 14C behaviour in Zr fuel clad wastes under disposal conditions (See Deliverable D3.20 for further information and references)
The presented study focuses on Zircaloys for which 14C is found either in the alloying 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.
A literature review was initially carried out to establish the State of the Art. The corrosion behaviour of zirconium alloys was reported in the review highlighting the high resistance to uniform corrosion at low or moderate temperatures. Various studies have shown that the uniform corrosion rates of zirconium alloys are very low in anaerobic neutral or alkaline waters at low temperature with an envelope value of 20 nm.y-1. The zirconia solubility remains very low for carbonate concentrations lower than 10-2 M. The susceptibility to localised corrosion (pitting, crevice corrosion and stress corrosion cracking) is negligible in anaerobic groundwaters. During the corrosion of Zircaloy, there is a possibility of a mechanism in which 14C is not released immediately by corrosion but is incorporated into the oxide film and then released by diffusion or during the zirconia dissolution. It is supported by the fact that measured 14C specific concentrations in zirconia oxide layers are about twice of that of Zircaloy metal after irradiation in a reactor. CAST Final report on 14C behaviour in Zr fuel clad wastes under disposal conditions (D3.20). In the CAST project, various irradiated and unirradiated Zircaloys (Zr, Zr-2, Zr-4 and M5TM) have been studied. Particular attention was brought to characterise the materials in terms of microstructure, composition and oxide layer by means of optical microscopy, scanning electron microscopy (SEM), transmission electron microscopy (TEM) and X-ray diffraction (XRD). The characterisations were carried out before and after running the experiments. The total 14C inventory has been determined both experimentally and by calculations. The results seem to be in good agreement. The vast majority of 14C inventory (90%) is released as gaseous organic compounds during dissolution of Zircaloy into the gas phase.
Leaching experiments were conducted in alkaline media (NaOH and Ca(OH)2) under anoxic conditions at room temperature for several time durations ranging from 14 days to 6.5 years.
Considerable effort was put to develop analytical techniques able to measure 14C in low concentrations, such as those expected in deep geological repository conditions. The liquid scintillation counting (LSC) was widely used to determine the total 14C released in alkaline environment as well as the inorganic / organic partition. The measurements of 14C in leaching solution were performed after decontamination from other high activity level radionuclides (Cs-137, Co-60, Sb-125, Ru-106/Rh-106) by ion exchange resin. Speciation and quantification of dissolved 14C were performed by using separation technique (collection of fractions by ion chromatography) and analysis by liquid scintillation counting method. In addition, accelerator mass spectroscopy (AMS) was used to quantify organic molecules in very low concentrations. Carboxylic acids such as oxalate, formate and propionate could be determined.
The gas samples were analysed by a gas chromatograph (GC). The results revealed the production of methane, ethene and CO2.
The results highlight a fast release fraction of 14C as various experiments showed a relatively constant concentration of 14C released in static alkaline solution.
Further, corrosion measurements were performed on both, unirradiated and irradiated Zircaloys by using hydrogen measurements and electrochemical measurements such as linear polarisation resistance (LPR). Overall, the results showed that the corrosion rates decreased with time. Electrochemical measurements enhanced discrepancies of the results while hydrogen measurements seem to be the most reliable technique to measure the corrosion rate.
From a safety assessment point of view, the instant release fraction (IRF) was determined on irradiated Zircaloy-2 based on inventory measurements. The results showed that the 14C inventory in the oxide was around 7.5%, which is below the 20% commonly used in safety case assessments.
All in all, longer time experiments should be conducted in order to obtain steady state conditions and confirm the current results.

Work Package 4: Spent Ion-Exchange Resins 14C Source Term and Leaching (See Deliverable D4.9 for further information and references)
It is confirmed that the major part of the 14C source term in SIERs is in inorganic form: more than 90% in BWR samples, and ca. 70-80% in PWR samples; the organic forms are mainly short chained organic acids (formate, acetate); a small amount of gaseous release can be obtained, in inorganic form (CO2); a strong alkaline solution is necessary to leach 14C from SIERs; no leaching from the cemented SIERs samples was observed.
Some general comments can be made:
• It is confirmed that the major part of 14C source term in SIERs is in inorganic form, i.e. as part of the anions from the carbonate system: more than 90% for SIERs from BWR, and more than 70% from PWR;
• The organic fraction of the 14C source term is mainly found as short chained organic acids, i.e. formate or acetate;
• A gaseous release has only been observed under alkaline conditions, and in the form of inorganic 14C;
• The leaching of 14C from SIERs has only been observed under strongly alkaline conditions;
• The leaching of 14C from cemented SIERs has not been observed.
As noted in RIZZATO et al. [2017], it is nevertheless advised that a “systematic collection and investigation of SIERs from different water-cleaning circuits would favour a better understanding of the source term for 14C and could allow an optimization of the storage conditions in order to prevent uncontrolled release of 14C. This could also support the development of an effective treatment process in order to transfer SIERs into more robust waste-forms for long-term disposal”. It is also recalled that the “mechanisms evidenced in the present work, however, have been obtained for unconditioned SIERs. National strategies for the management of SIERs are different and conditioned/treated SIERs may not reflect the same behaviour”.

Work Package 5: Irradiated Graphite (See Deliverable D5.19 for further information and references)
Work Package 5 involves the contribution of a number of organisations in relation to irradiated graphite (i-graphite) studies (“Irradiated graphite” is referred to in some national programmes as “nuclear graphite”). This report summarises their progress in CAST on an organisation-by-organisation basis. Note that the work undertaken by each organisation is affected by the national position in relation to the management of i-graphite, which in some cases involves treatment activities. In this Executive Summary, the top-level output of each organisation in CAST Work Package 5 is presented.
1 CNRS/IN2P3 (IPNL) participates in Task 5.2, named “Characterisation of the 14C inventory in i-graphites” of Work Package 5. This task aims at characterising the 14C inventory in i-graphites because such information is essential in understanding the release. CNRS/IN2P3 (IPNL) develops an indirect approach based on ion-irradiation of 13C-implanted virgin graphites (HOPG model graphite). 13C implantation is used to simulate 14C displaced from its original structural site through nuclear recoil. Ion irradiation is used to simulate neutron irradiation and study the effects of both graphite structure modification and irradiation on the migration of 14C. The results are presented in Deliverable D 5.18.
2 During the first three years of the CAST Project, LEI concentrated on the performance of the Task 5.1 – “Review of CARBOWASTE and other relevant R&D activities to establish the current understanding of inventory and release of 14C from i-graphites” and Task 5.2 – “Characterisation of the 14C inventory in i-graphites”. Task 5.5 “Data interpretation and synthesis – final report” is a final project year activity. For the Task 5.1, LEI reviewed the outcome of CARBOWASTE Project in the national context and based on that, provided input for deliverable D5.5 “Review of current understanding of inventory and release of 14C from irradiated graphites”. For the Task 5.2, LEI modelled the 14C inventory in a RBMK-1500 reactor core using new developed models and experimental data and prepared deliverable D5.17 “Report on modelling of 14C inventory in RBMK reactor core”. The Task 5.5 “Data interpretation and synthesis – final report” is the last Task of WP5 that summarises and synthesises in a final report the work undertaken in the previous Tasks by all participants. The outcome of this Task is this deliverable D5.19 “Final report on results from WP5”.
3 RATEN ICN was involved in Task 5.1 and Task 5.3. For the Task 5.1, RATEN ICN reviewed the outcome of CARBOWASTE project in the Romanian context as well as the 14C content and speciation and their correlation with impurity content and irradiation history in the MTR i-graphite. The results of the review of the current knowledge achieved in the understanding of the speciation of 14C from irradiated graphite used in research reactors are included in D5.5 “Review of Current Understanding of Inventory and Release of 14C from Irradiated Graphite”.
For Task 5.3, RATEN ICN carried out leaching tests, in aerobic and anaerobic conditions, to assess the 14C release as dissolved species in alkaline solution simulating the cementitious environment. The irradiated graphite available for the leaching tests is originating from a brick dismantled in 2000 from the thermal column of the TRIGA research reactor in operation at RATEN ICN. The total 14C measurement in the powder graphite samples taken during mechanical cutting of the specimens used in the leaching tests was achieved by combustion in oxygen reach atmosphere. Also, gamma emitting radionuclides were measured by gamma spectrometry on the same powder graphite samples. For the measurement of inorganic and organic fraction of the 14C released in leachate solutions an analytical method based on acid stripping and wet oxidation was applied. The 14C measurement were carried out by liquid scintillation counting (LSC) using a Tri-Carb® analyser Model 3110 TR. The experimental results obtained from the leaching tests both in aerobic and anaerobic conditions confirm the low 14C release in alkaline environment. Less than 2% from the total 14C inventory in the specimens that were subjected to the leaching tests was released as dissolved species. Under aerobic conditions, the 14C released was found to be predominantly in inorganic forms (around 68% from the total 14C released), while in anaerobic conditions, the 14C is released mainly as organic species (around 65% from the total 14C released). These results are presented in D5.10 “Final report on 14C release and inorganic/organic ratio in leachates from TRIGA irradiated graphite”
4 During the CAST project, Andra and EDF shared some results from studies on French i-graphite. The first step was to review the existing leaching data on 14C in French i graphites. Deliverable D5.1 was devoted to the synthesis of the leaching rates of 14C. In deliverable D5.8, the second step was to study the speciation of the released 14C in the leaching liquor and in the gas phase.
5 ENEA participated in Tasks 5.4 “Exfoliation of irradiated nuclear graphite by treatment with organic solvent assisted by ultrasound”, within Work Package 5 (WP5) of the CAST Project. The behaviour of nuclear graphite (NG) through the exfoliation process has been studied, carried out with different organic solvents to remove primarily radiocarbon and estimate the “Removal Efficiency” (as percentage of the recovered activities after treatment in comparison to the original values) of the used solvents, in order to propose a reliable approach to produce graphite for recycling or/and safety disposed as L&ILW.
The exfoliation method is a reliable method to remove some 14C from nuclear graphite, paving the way for the production of graphite for recycling and/or safe disposal. To support this method, a lot of evidence has been collected.
It has been demonstrated that the ratio powder/solvent has a crucial position in the optimization of the whole procedure. The recovery of exfoliated graphite increases when the ratio powder/solvent is low: to maximize the effect of the solvent it could be very useful to operate with a recovery and recycling system, also due to the fact that the solution produced from the exfoliation of nuclear graphite is highly radioactive.
The extracting abilities of each solvent are quite comparable. Since the partially removed activity is independent of the yield degree, these results can be interpreted on the basis that not all the 14C species in the irradiated graphite are the same and only the ones not chemically bonded or in those forms that have some chemical or physical-chemical affinity with the extracting media are likely to be removed.
These first results are promising enough to promote further investigations in this direction. Furthermore, there is no doubt that the basic aqueous mixture is more efficient in the removal of 14C respect to the organic solvents. NaOH increases the production of carbonate (highly soluble in water) confirming that the next experiments have to be designed following this evidence, focusing on efforts to develop a “total green” procedure.
6 During three years of CAST project FZJ participated in the Tasks 5.1-5.3 and 5.5 of WP5, providing contributions to a number of reports and deliverables. Within the first year FZJ contributed to the “Review of CARBOWASTE and other relevant R&D activities to establish the current understanding of inventory and release of 14C from i-graphites” (D5.5) and “Characterisation of the 14C inventory in i-graphites” (D5.6). The relevant to the project information on different German i-graphite (FRJ-1, FRJ-2, AVR, HTHR-300, RFR) was summarized and provided to the CAST-community. Within the Task 5.3 FZJ comprehensively characterized RFR i-graphite and investigated the leaching behaviour of RFR i-graphite, providing for contributions to the “Definition of the scientific scope of leaching experiments” (D5.4), Annual Progress Reports (D 5.6 and D5.9), as well as to the “Final evaluation of leach rates of treated and untreated i-graphite from the Rossendorf Research reactor” (D5.12). The outcome of the Task 5.5 is the current “Final report on results from WP5” (D5.19).
7 Throughout the CAST project CIEMAT has been involved in four of the five main tasks of the WP5: Graphite. These tasks are Task 5.1 “Review of CARBOWASTE and other relevant R&D activities to establish the current understanding of inventory and release of 14C from i-graphites”, Task 5.3 “Measurement of release of 14C inventory from i-graphites”, Task 5.4 “New waste forms and 14C decontamination techniques for i-graphites” and Task 5.5 “Data interpretation and synthesis – final report”. Regarding the Tasks 5.1 and 5.5, CIEMAT provided its contribution to the deliverables D5.4 “Definition of a recommended scientific scope of leaching experiments and harmonized leaching parameters”, D5.5 “WP5 Review of Current Understanding of Inventory and Release of 14C from Irradiated Graphite” and the current report (D5.19), all generated as an output of these tasks.
For the Task 5.3, CIEMAT developed both analytical methods and protocols to measure the release of 14C from irradiated graphite samples and the speciation in the aqueous and gaseous phase. The conclusions of these works were depicted in the report D5.15 “WP5 CIEMAT Final Report on 14C Leaching from Vandellós I Graphite”.
Finally, for Task 5.4, CIEMAT manufactured a new impermeable graphite matrix denominated (IGM), and investigated the behaviour of this material (corrosion progress and leaching rates of radionuclides) in contact with two types of aqueous media, deionized water and granitic - bentonite water. An additional research line consisted on the validation of the 14C decontamination procedure based on the thermal treatment of the Vandellós I NPP graphite. The outcome of these works was reported in the document D5.13 “Vandellós I graphite compilation report: thermal decontamination and IGM waste form results”.
8 During the first four years of the CAST Project, the activities carried out by IFIN-HH were focused on three directions: Task 5.1 – “Review of CARBOWASTE and other relevant R&D activities to establish the current understanding of inventory and release of 14C from i-graphite”, Task 5.2 – “Characterization of the 14C inventory in i-graphites” and Task 5.3 “Measurement of release of 14C inventory from i-graphites”. For the Task 5.1, IFIN-HH reviewed the outcome of CARBOWASTE Project in the national context and based on that, provided input for deliverable D5.2 “Review of current understanding of inventory and release of 14C from irradiated graphites”. For the Task 5.2, IFIN-HH explored the use of accelerator mass spectrometry to measure 14C distributions in i-graphite and beta imaging to determine the distribution of 14C and H-3 on graphite surfaces and prepared deliverable D5.7 “Report on 14C distribution in irradiated graphite from research reactor VVR-S using accelerator mass spectrometry and beta imaging”. In the Task 5.3 “Measurement of release of 14C inventory from i-graphites” IFIN-HH developed a separation technique, based on silica gel columns and oxidation at high temperature over a CuO catalyst be decoupled with LCS device (TRICARB 2800 TR PerkinElmer) in order to measure the release rate of 14C into gas and solution phase from irradiated graphite (intact and crushed samples from VVR-S reactor and prepared deliverable D5.14 “Release of 14C from irradiated VVR-S graphite to solution and gas phase (Task 5.3)”. IFIN-HH also contributed to the last Task of WP5 (Task 5.5) that summarizes and synthesises in a final report the work undertaken in the previous Tasks by all participants. The outcome of this Task is this deliverable D5.19 “Final report on results from WP5”.
9 Work reported by Radioactive Waste Management (RWM) considers the release of gaseous 14C from irradiated graphite samples from Oldbury Magnox power station. Key conclusions from the study are that under baseline conditions, the predominant 14C release was to the solution phase, with about 0.07% of the 14C inventory being released into solution in one year and with about 1% of the released 14C being released to the gas phase. An initial rapid release of ~3 Bq of 14C to the gas phase was observed in the first week, and subsequently the rates of release decreased with time but detectable quantities of 14C were found in all gas samples. For the final cumulative sampling period, a release of ~2 Bq was measured. Gaseous 14C was predominantly in the form of hydrocarbons and other volatile organic compounds and CO, and less than 2% of the gas-phase release was in the form of 14CO2. The differences between the fractional releases of 14C from Oldbury graphite and that released from BEP0 graphite – a subject of previous study - are likely to be due to differences between the original graphites, their irradiation histories, operating temperatures and coolant gases; the results of this study do not allow determination of whether any one of these differences is a dominant effect although a wider comparison of irradiated graphites might.
Over the course of the EC CAST project, RWM undertook its role as Work Package 5 leader.
10 For Ukraine, the main radiocarbon source is irradiated graphite from the Chernobyl Nuclear Power Plant. The ChNPP is a decommissioned nuclear power station about 14 km northwest of the city of Chernobyl, and 110 km north of Kyiv (Kiev). The ChNPP had four RBMK reactor units. The commissioning of the first reactor in 1977 was followed by reactor No. 2 (1978), No. 3 (1981), and No.4 (1983). Reactors No.3 and 4 were second generation units, whereas Nos.1 and 2 were first-generation units. RBMK is an acronym for "High Power Channel-type Reactor" of a class of graphite-moderated nuclear power reactor with individual fuel channels that uses ordinary water as its coolant and graphite as its moderator. The combination of graphite moderator and water coolant is found in no other type of nuclear reactor.
The approved «ChNPP Decommissioning Program» establishes a preferred decommissioning strategy.
Work undertaken by IEG NASU in CAST presented information on graphite characterisation as is being undertaken as part of the ChNPP Decommissioning Program. Work considering chemical decontamination of graphite is reported, as is the overall graphite waste management approach.
Work of IEG NASU in the EC CAST project targeted provision of input on the use of graphite in the RBMK reactor, particularly the decommissioning of each category, classification, volume, weight and activity (the inventory of i-graphite in Ukraine is not finished: in compliance with the requirements of the regulatory documents, the composition and activities of radionuclides accumulated in structural materials and structures during the operation of the NPP power unit must be evaluated before its removal from service.). IEG NASU reviewed characterisation data on the speciation of 14C in RBMK reactor, i-graphites and relevant information from decommissioning projects in Ukraine. IEG NASU is also involved in the review of the outcome of the IAEA project investigating conversion of i-graphite from decommissioning of Chernobyl NPP into a stable waste form.

Work Package 6: Synthesis report CAST outcomes in the context of the safety case: Improved understanding of 14C release from CAST –WP2 to WP5 input to WP6 (See Deliverable D6.4 for further information and references)
14C is present in important amounts in the inventory of many waste families, and particularly in activated steel, Zircaloy (Zy), graphite and spent ion exchange resins (SIERs). The knowledge regarding the chemical form and the release mechanism of 14C from these wastes in disposal conditions is limited. Conservative treatments must thus be adopted in safety assessments to cope with these uncertainties, giving rise to overestimated radiological impacts. 14C shows contrasting behaviours in safety calculations depending on whether it is in inorganic or organic form. As carbonate, 14C shows excellent retention characteristics in cementitious and clay environments due to isotopic exchanges, whereas it is much more mobile and possibly in gaseous phase as an organic species. The primary focus of CAST was thus to discriminate experimentally between these two different forms and to strive for a more precise characterisation of the speciation of the 14C bearing compounds released from the wastes investigated in this project.
The measurements of 14C speciation released from steel show that both organic and inorganic compounds are present in the liquid phase [MIBUS et al., 2018 (CAST D.2.18)]. Methane and minor contributions of CO are found in the gas phase. However, applying this speciation to long-term releases of 14C in disposal conditions is debatable. The oxygenated species measured in experimental conditions might in fact result from the radiolysis induced by the activated materials. It could thus be expected that the 14C speciation will shift to reduced compounds, such as – gaseous – hydrocarbons, when the radiolysis becomes ineffective in the disposal system. 14C from Zircaloy shows the same behaviour in terms of speciation: The liquid phase is shared between inorganic and small oxygenated organic compounds. Methane, ethene and CO2 were mainly detected in the gas phase. Precise distribution as an input to safety assessment is still challenging at this stage, nevertheless, it can be concluded that the organic form of 14C released from Zircaloy and steel is present in non-negligible fractions.
For Zircaloy and steel, exhaustive literature reviews and experimental studies on their corrosion rates were carried out in CAST to bound the 14C source term. Steel corrosion rates in alkaline conditions are very low because of the presence of a passivation layer [SWANTON et al., 2015 (CAST D2.1); MIBUS et al., 2018 (CAST D.2.18)]. Many experimental studies have shown that uniform corrosion rates for carbon steel in anoxic, alkaline conditions are below 0.1 μm/year. Recently, an increasing number of studies indicate upper values in the range of few tens of nm/year. Uniform corrosion rates for stainless steel are very low. In anoxic, alkaline conditions, recent studies reported measured values below 0.01 μm/year. The corrosion rates of stainless and mild steels are higher in neutral conditions than in alkaline conditions. This might affect the 14C release in a pessimistic scenario where the alkalinity of the near field decreases i.e., because of ageing of the cementitious environment. However, the radiological impact remains limited since the pH decreases significantly only in the long-term when most 14C has decayed. The experimental studies indicate an early, fast release of 14C from steel between a negligible fraction up to a few percents. There is no consensus on how to abstract these observations in safety assessment.
The experimental works conducted on the corrosion of Zircaloy in CAST tends to confirm the data reported in the literature with corrosion rates in the order of a few nm/year at the most at low temperature, in alkaline or neutral conditions [GRAS, 2014 (CAST D3.1); NECIB et al., 2018 (CAST D3.20)]. CAST allowed progress in the knowledge of the corrosion mechanism of Zircaloy. Should the corrosion regime change in disposal conditions (transition to pseudo-linear kinetics), it is not expected to lead to higher rates. Further, the CAST results on Zircaloy confirm the hypotheses formulated fifteen years ago by the Japanese program of a mechanism in which 14C is not released immediately by Zircaloy corrosion but is retained inside the oxide film. Indeed, the total leached fraction of 14C from long-term Japanese experiments of several years on PWR and BWR cladding samples is less than 0.1%. The 14C released in this experiment seems to originate from the oxide layer [SAKURAGI et al., 2018 (CAST Final Symposium)]. These experimental results suggest that the 20% instant release fraction (IRF) used traditionally in safety assessment is over-conservative. Unfortunately, there are not enough data and currently no consensus over the release mechanism of 14C from the oxide layer to abstract these very low fractions in quantitative safety assessments. In addition, the influence of hydrides on the corrosion behaviour on the long term in disposal conditions remains uncertain. Last, let’s note the reviewing work performed in CAST over the Zircaloy inventory in the claddings that reduced the uncertainties on the concentration of the nitrogen impurity [GRAS, 2014 (CAST D3.1); CAPOUET et al., 2017 (CAST D6.2)]. However, there remains a certain uncertainty on the 14C inventory in reprocessed waste (vitrified and compacted waste). The assumptions regarding the carry-over fractions of 14C inventory from the different components of the assemblies need to be consolidated. Also, accounting for an accessible 14C in the oxide layer in compacted waste (after acid treatment) is still a matter of debate [CAPOUET et al, 2017 (CAST D6.2)].
The study of 14C in irradiated graphite in CAST lies in the continuity of CARBOWASTE. A certain number of outcomes were highlighted in [TOULHOAT et al., 2018 (CAST D5.19)]. First, regarding the release rate, a substantial fraction of the 14C in irradiated graphite is not releasable. Some 14C will initially be released rapidly, and some will be released more slowly at a rate that decreases over time. 99% of the 14C released is in the inorganic form. 14C can be released to both the gas and aqueous phases. A number of different species, including organic species (e.g., CH4), CO2 and CO may exist in the gas phase. For high pH conditions, the proportion released as gaseous carbon dioxide is small in comparison to the fractions released as carbon monoxide and methane. CAST provided a consensual parameterisation of the 14C source term in irradiated graphite as basis for safety assessment. However these data should be considered with care as the release rate and speciation of 14C is function of the graphite type used in different reactors and the disposal concept and conditions.
SIERs is a very heterogeneous source. The range of activity of SIERs depends on specific factors such as reactor and circuit type, history of the physico-chemistry in the fluid as well as pre-disposal storage conditions and conditioning processes of the resins. Likewise, 14C speciation is expected to be influenced by these factors. In the case of BWR more than 90% of 14C was found under the form of inorganic carbon. For PWR, the situation is more contrasted. CANDU reactors seem to induce a major part of inorganic 14C, whilst for PWR around 20% of 14C was obtained [REILLER, 2018 (CAST D4.9)]. The speciation of the organic fraction suggests formic acid as the main organic form in SIERs. The conditioning matrix of SIERs (epoxy and cement) is assigned a safety function of water ingress limitation and possibly retardation in safety assessment [CAPOUET et al., 2017 (CAST D6.2)]. Experimental studies brought to light the lability of 14C in -unconditioned- SIERs during predisposal processing: The presence of atmospheric air during storage, temperature increase, transient decrease in the pH upon contact with alkaline solutions as well as drying procedures of the SIERs seems to cause a release of inorganic carbon [RIZZATO et al., 2017 (CAST D4.8); REILLER, 2018 (CAST D4.9)]. SIERs are a telling example illustrating the strong dependency between predisposal and long-term disposal management strategies.
Upscaling to geological disposal systems
Due to its relatively short half-life, the 14C radiological released from the waste through the host rock is sensitive to the release and migration processes in the disposal system. Clay disposal systems provide an excellent performance regarding 14C, provided transport times in the disposal system are of the order of a few tens of thousands of years. The 14C activity releases from the geological barrier (whatever its chemical form) are barely sensitive to the instant release fractions (IRF) up to 20% due to the spreading effect of the diffusive transport [HENOCQ et al. 2018 (CAST D6.3)]. Sensitivity studies show that, in the present state of knowledge reported by PSI in CAST, the possible uptake capacity of the 14C bearing organic compounds identified in CAST is to weak to influence the 14C transport [CAPOUET et al., 2017 (CAST D6.2)]. The impact of organic 14C might become more relevant in scenarios where diffusion through the geological barrier is cut short, as could occur for example in the case of a scenario that considers transport of groundwater and dissolved species by advection. The impact of this scenario is very dependent on the uncertainties pertaining to corrosion rates, amount of metals and their specific surfaces. Reducing these uncertainties would make it possible to better estimate the source term of both the hydrogen carrier and 14C. Different design strategies can also be applied to limit the impact of both the hydrogen pressurisation and the advective transport of 14C [HENOCQ et al., 2018 (CAST D6.3)].
Locating the repository caverns in crystalline rocks away from any major fracture zones at a sufficient depth limits the groundwater flow through and in the vicinity of the repository caverns. This provides favourable near-field conditions for the engineered barrier system, limits the radionuclide transport and isolates the waste from the biosphere. In crystalline rocks, any open fractures can provide pathways for both gas and aqueous transport. The release and migration of 14C in organic gaseous form is expected to occur at a rate comparable to the migration of organic 14C dissolved in groundwater. Sensitivity analyses carried out in CAST of 14C releases from a repository located in a crystalline rock indicate a strong impact of the near field processes, i.e. groundwater flow rates and sorption on the release rates. Consequently, the transport and retardation properties of the aging cementitious environment are critical. Radiological impact in crystalline rock is more sensitive to IRF and (potential) low distribution coefficients (assigned to cement and host formation) than in clay. Reducing the uncertainties related to the metal corrosion rate as well as a good knowledge of the cementitious evolution is of primary importance for crystalline systems [HENOCQ et al., 2018 (CAST D6.3)].
A repository for radioactive waste in a salt formation is characterised by mostly dry conditions. Therefore, radionuclide transport occurs dominantly though the gas phase. The convergence of the backfill starts as soon as the disposal is closed. This process can be the driver of an advective transport of gases though the EBS up to the biosphere. A potential release of 14C from the repository in salt depends thus on the amount of gaseous 14C made available in the early few hundreds of years after disposal closure, due to corrosion by water brought during the operational period or due to initial canister failure. The experimental conditions of CAST (saturated and alkaline) are not directly representative of the conditions prevailing in a salt disposal (unsaturated & high saline brines). Water being a limiting factor, the impact in salt is very sensitive to the gaseous IRF. Although conditions are different, the literature (First Nuclides) relevant for salt system is in line with CAST. It shows increasing indications that the gaseous release of accessible 14C from both the Zircaloy (oxide layer), the spent fuel rod and the steel are relatively low (1% all together). The highest priority for salt disposal systems is to reduce the uncertainty on the release behaviour of gaseous 14C. This is mainly related to three questions:
1/ What is the percentage of 14C which can be released in volatile form?
2/ What is the temporal distribution of this release?
3/ Is water necessary to transfer 14C into volatile form or does this occur in dry conditions? [HENOCQ et al., 2018 (CAST D6.3)].
Key messages to safety case
In conclusion, the experimental studies of CAST have confirmed the release of a non-negligible fraction of 14C organic compounds from steel, Zircaloy and SIERs in alkaline and anoxic conditions. Regarding Zircaloy and steels, hydrocarbons and carbon monoxide were found in the gas phase whereas the aqueous phase contained small oxygenated organic compounds. The mechanism of formation of these organics remain uncertain, in particular the source of oxygen. Although the organic nature of 14C products generated from steel and Zircaloy corrosion has been confirmed, long-term generation of 14C in disposal conditions might give a different picture with respect to its organic speciation and compound distribution. Consequently, conservative treatment still applies in safety assessment regarding specific organic speciation. CAST gave the opportunity to reinforce the understanding of the corrosion mechanisms of these metals, in alkaline, anoxic conditions. As a result, the confidence that these corrosion mechanisms will remain generally unchanged in the long term (within a certain Eh/pH window of the near field) has increased. Also, the interplay of the oxide layer in the 14C release mechanism of Zircaloy is now acknowledged. The literature review carried out in CAST confirms the low corrosion rates for these metals as well as the general trend to even lower rates as observed in more recent studies. CAST emphasized the heterogeneous character of irradiated graphite and spent ions exchange resins. The relative importance in the safety case of 14C (aqueous) versus 14C (gaseous) for these wastes vary by disposal concept, predisposal activities, and operational conditions. Applying the results determined from few specific samples to broad inventories of waste with various operational and predisposal histories must be done with caution. This generalisation process might bring a certain level of uncertainty to be accounted for in safety case. Safety assessment studies carried out in CAST highlighted the critical influence of the chemical and physical evolution of the cementitious environment on different aspects of the 14C source term (e.g., corrosion rates, 14C release rates), but also on more global aspects pertaining to the confinement properties of a geological disposal (e.g., fate of the hydrogen produced by corrosion, near field hydraulic properties). 14C in the form of a mobile organic compound will give a more relevant radiological impact than if considered in the inorganic form, and this particularly in rapid transport scenarios. Reducing the uncertainty on 14C speciation shifts the conservatism introduced in safety assessment of 14C release to the corrosion and transport rates.

Work Package 7 Dissemination: Final overview of CAST (See Deliverable D7.23 for further information and references)
This overview intends to summarize the results and achievements of CAST and their implications for building confidence in the safety of geological disposal. The half-life of 14C is 5730 years. In many disposal concepts for High Level Waste (HLW), there is an engineered containment period up to several hundred thousand years. The activity concentration of 14C in waste is significantly reduced in this containment period. Investigation of the 14C Source Term is therefore of more interest for the Low and Intermediate Level Waste (LILW) than HLW for many countries. Experiments to measure the release of 14C and 14C speciation have therefore been mainly performed at room temperature. The CAST project focused on the release of 14C as dissolved and gaseous species from waste. 14C can be released as an organic species i.e. carboxylic acids, alcohols, aldehydes, alkenes and alkanes or as an inorganic species i.e. carbonate, bicarbonate and carbon dioxide. All species need to be measured, in order to determine the main released 14C species.
The measurement of released 14C species for the neutron irradiated materials is challenging due to its low chemical amount in waste and low corrosion rates in cementitious materials. Typical 14C activity concentrations in neutron irradiated steel from the core of a nuclear plant are in the order of 1E5 Bq per gram solid matter and for neutron irradiated Zircaloy 1E4 Bq per gram solid matter i.e. a few ppm in neutron irradiated steel or tenths of ppm in neutron irradiated Zircaloy. The speciation of released 14C species could not be quantified within CAST. These materials contain non-radioactive carbon: for example in steel, the non-radioactive carbon content is three orders in larger magnitude than the 14C content. In CAST, significant progress has been made to quantify the speciation of released non-radioactive carbon from steel. The major 14C species were dissociated anionic compounds from carboxylic acids in the case when these dissociated anionic compounds and gaseous organic carbon were measured; gaseous organic carbon was negligible compared to dissolved organic compounds. An analytical strategy for the quantification of released alcohols and aldehydes has been developed. For neutron irradiated steel, carbon is released as dissolved non-gaseous non-ionic compounds i.e. alcohols and aldehydes. These compounds cannot be retarded in the cementitious materials by sorption and ion exchange due to lack of electrostatic interactions.
The presence of hydrogen can influence the fate of released carbon species; the 14C released as non-ionic compounds such as alcohols and aldehydes can be further reduced with hydrogen to gaseous organic carbon. Significant confidence has been raised that more than 90% of the hydrogen generated during anaerobic corrosion is picked-up during corrosion of Zircaloy at disposal conditions. No hydrogen is formed during degradation of graphite and ion exchange resins under disposal conditions, but hydrogen can be formed by radiolysis of concrete pore water by radiation from neutron irradiated graphite and spent ion exchange resins. These radiolytic hydrogen generation rates are usually neglected. Consequently, it is only for neutron irradiated steel that the potential reduction to gaseous organic carbon needs to be taken into account. Carboxylic acids are expected only to occur as dissociated anionic compounds at disposal; their presence hinders the reduction to gaseous alkenes and alkanes such as ethene, methane and ethane with hydrogen. Only in acidic media for example in acid digestion tests, this potential reduction of carboxylic acids is expected to be possible.
The OH radicals formed by radiolysis of concrete pore water can also have an influence on the fate of released organic carbon. These radicals can oxidise alcohols and aldehydes into inorganic carbon. This process may have occurred during the experimental investigations in CAST in which neutron irradiated steel and neutron irradiated Zircaloy were exposed to portlandite solutions. Gases released from these solutions were measured to be organic gaseous carbon as well as carbon dioxide. Radiolysis may be the reason why both reduced and oxidised hydrocarbons can exist simultaneously in solutions in contact with corroding metals. If the measured CO2 is presumed not to be a contaminant, the carbon dioxide may have been formed by oxidation of dissolved organic compounds with OH radicals. Also, in alkaline solutions with spent ion exchange resins, 14CO2 has been measured. This induced oxidation of dissolved organic compounds is positive for a potential retardation in cementitious materials, since any CO2 released from waste in cementitious materials will form an unstable carbonic acid that is dissociated into bicarbonate and carbonate. The high calcium content in concrete pore water and low solubility will cause precipitation. Another retention mechanism of carbonate is sorption on positively charged cementitious minerals.
Dissolved organic carbon compounds have been measured to be the major 14C species released from neutron irradiated steel, neutron irradiated Zircaloy and neutron irradiated graphite. Distinction between carboxylic acids, alcohols and aldehydes is to be made. The release of non-ionic carbon compounds such as alcohols and aldehydes is therefore conservatively assumed to be the main 14C species. The extent to which these species can be assumed to be retarded in engineered and natural barriers is uncertain, and in the case of natural barriers depends on the host rock properties. These neutron irradiated materials are also known to have very small mass loss rates, so the potential 14C fluxes to air from these three waste types as a consequence of their deep geological disposal is expected to be negligible compared to the flux of natural 14C emanating from soil.
The main origin of 14C in neutron irradiated steel and Zircaloy is nitrogen. Measured and calculated 14C contents are of the same order in magnitude. For neutron irradiated graphite, it depends on the chemical and temperature conditions whether nitrogen is the main contributor to 14C in waste. Nitrogen-activated 14C forms a loosely bound 14C gaseous species that can e.g. be released during crushing of neutron irradiated graphite.
Only measured 14C contents in resins that have been prepared for conditioning are representative of behaviour likely in relation to a disposal scenario, due to the high impact of the storage and drying conditions on the contained 14C content. Knowledge of the
contained inorganic and organic carbon content is necessary for safety assessments when organic carbon compounds are assumed to be non-retarded species. Conservatively, the speciation of 14C contained by resins is assumed to be released instantaneously. Spent ion exchange resins can only contain anionic 14C compounds. This containment limits the presence of organic carbon in ion exchange resins to dissociated anionic compounds of carboxylic acids such as oxalate and acetate; the only possible contained inorganic compounds are bicarbonate and carbonate. Attempts to measure release of 14C in alkaline media have been made but 14C has not been measured to be released in cementitious pore water. Inorganic carbon is assumed to be retarded in cementitious materials due to the precipitation with calcite i.e. fixation by high calcium amount in the concrete pore water chemistry. This retention mechanism also takes place for oxalate. Sorption of positively-charged cementitious minerals is another retention mechanism and in CAST the literature for sorption of acetate has been found. The last evaluated retention mechanism is ion exchange; formate is exchanged with sulphate in ettringite. Consequently, there is no reason in cementitious materials only to assume retention for inorganic carbon and not for organic carbon during disposal of spent ion exchange resins when processed with cementitious materials. The amount of 14C that is released and not decayed within the waste package may be negligible.

Advisory Group reports on each of the above six synthesis reports are available.

Potential Impact:
CAST has ensured ensure robust and efficient cross-fertilisation of knowledge, understanding, information and experience amongst many countries and organisations, both within and outside Europe. It will allow opportunities for international progress on the integral issue of 14C generation to be reflected in national programmes, and for the partners' progress to be highlighted internationally. CAST has provided invaluable experimental data that can then be applied by the national WMOs to their programmes. In addition CAST has put put these experimental results into the context of national safety assessments.

WP 2, 3 and 4 have similar impact in providing significant data and understanding on 14C generation from steels, Zircaloys and ion-exchange resins respectively. WP 5 has consolidated and furthered the current understanding of 14C release from irradiated graphite. WP 6 will have perhaps the largest and most significant impact for the end users of the data as, drawing on the results from WPs 2 to 5, it will lead to a greater understanding for WMOs of the impact of 14C in safety assessments for disposal facilities. WP7 has been established with a specific dissemination objective and significant effort is being made to disseminate the project results most effectively and widely. This will include presentations at international conferences and other meetings, publication in journals and conference proceedings, and through two workshops to be held over the duration of CAST. The first workshop outlined the initial findings from CAST, and allow interested parties to familiarise themselves with the proposed research, and to ask questions of WP representatives. The second workshop presented the overall results and findings of CAST and discussed these with the target groups.

Training is a key aspect of CAST by the engagement of PhD students, post-doctoral scientists and the ongoing development of existing staff. Two training courses will be provided for junior participants. The first provided experience in radioanalytical techniques and sample handling; the second will provide understanding of 14C generation and wastes, and experience in transport modelling. It is also be possible for early-career researchers to exchange with other partners.

Dissemination of CAST output:
Work Package 7 ensures the dissemination of output from CAST; please see associated deliverables.
A special edition of the journal Radiocarbon is in preparation, which will contain a number of papers covering various aspects of CAST output. As at 2nd July 2018, the papers submitted are titled as follows:
1. C-14 content in CANDU spent ion exchange resins;
2. Carbon-14 release from TRIGA irradiated graphite;
3. C-14 content in CANDU spent fuel claddings and its release under alkaline conditions;
4. Release and speciation of carbon from Zircaloy-4;
5. Carbon-14 release & appreciation from carbon steel;
6. Release of 14C and 3H from irradiated graphite of the thermal column of VVR S Reactor to solution phase;
7. Investigation of 14C migration from RBMK-1500 reactor graphite disposed of in a potential geological repository in crystalline rocks in Lithuania;
8. C-14 release from steels under El Cabril Standard Leaching Test;
9. Preliminary analysis of gaseous radiocarbon behaviour in a geological repository hosted in salt rock;
10. 14C leaching and speciation studies on irradiated graphite from Vandellós 1 nuclear power plant;
11. 14C and other radionuclides in IGM waste form long term behaviour;
12. Identification of chemical form of stable carbon released from type 304L and 316L stainless-steel powders in alkaline and acidic solutions under low-oxygen conditions;
13. Estimation of 14C and other key radionuclides inventory in irradiated RBMK-1500 graphite based on limited number measurements and full 3D core modelling;
14. Investigation of Impurity of RBMK Graphite by Different Methods;
15. 14C release from steels under aerobic conditions;
16. Quantification of dissolved organic carbon-14-containing compounds by accelerator mass spectrometry in a corrosion experiment with irradiated steel;
17. 14C high concentration measurements with relevance for decommissioning of nuclear reactors.

Note that the deadline for paper submission has not yet passed, and hence other papers may become available. It is planned that the hard copy version of the special edition will be available in December 2018.

List of Websites:
http://www.projectcast.eu

Reported by

RADIOACTIVE WASTE MANAGEMENT LIMITED
United Kingdom

Subjects

Nuclear Fission
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