{"id":{"repo_id":"aachen","oai_identifier":"oai:publications.rwth-aachen.de:51873"},"canonical_url":"https://search.dev.ndltd.org/etd/aachen/oai:publications.rwth-aachen.de:51873","repository":{"repo_id":"aachen","name":"RWTH Aachen University","base_url":"https://publications.rwth-aachen.de/oai2d"},"display":{"title":"Graphite as radioactive waste : corrosion behaviour under final repository conditions and thermal treatment","abstract":"The present work deals with radioactive graphite management. Two different aspects were examined: 1. The corrosion behaviour of graphite under final repository conditions 2. The decontamination of graphite by thermal treatment Spent fuel from the German AVR and THTR high-temperature reactors is envisaged for direct disposal in a final repository. In this respect, corrosion of the graphite matrix of the spherical fuel elements and pyrocarbon coatings from coated fuel particles is a very important aspect for predicting the long-term behaviour of the radioactive waste. The graphite materials to be examined were isolated from unirradiated fuel elements, and their corrosion was investigated in different aqueous phases. The experiments were performed in deionised water, MgCl2-rich (brine-2) and NaCl-rich (brine-3) solutions at 90°C under argon, oxygen and air atmosphere. In order to investigate the influence of aqueous phase radiolysis, further experiments were performed in the presence of g-irradiation sources under argon atmosphere. It was established that the slow corrosion of carbonaceous materials in different aquatic phases in the absence of irradiation was caused by interaction with dissolved oxygen. Corrosion rates in aqueous solutions under pure oxygen and air atmosphere decrease in the order water > brine?3 > brine-2. Acceleration of graphite and pyrocarbon corrosion was observed in the presence of g-irradiation in brines while the opposite influence of the aqueous phases on the corrosion rates was additionally observed. The corrosion rates of graphite and pyrocarbon in irradiated brine-2 and brine-3 in argon atmosphere were two orders of magnitude higher than in pure oxygen atmosphere without irradiation. This may be related to the formation of highly oxidising chlorine species in brines, which react with carbon materials. In pure water, radiolysis did not influence the oxidation process significantly. However, calculating the expected lifetime of the graphite matrix in the repository on the basis of the corrosion rates determined is an extremely conservative consideration since the radiation dose rates used are significantly higher than could ever occur during final disposal of high-temperature reactor fuel elements. In the second part of the present work, thermal treatment of contaminated structural graphite from the high-temperature reactor core and thermal columns of the research reactors was investigated as a possible decontamination process. The main problem associated with direct disposal of contaminated graphite is its large volume. Reprocessing of the graphite based on graphite gasification offers the opportunity to separate the radionuclides from the main graphite mass, which could then be reused or disposed conventionally. The experiments were performed in an argon flow and steam in the temperature range of 870 – 1060°C. Comparison of the release rate ratios of the volatilised radionuclides 14C and 3H with the release of 12C showed that under all experimental conditions tritium and 14C were released faster than the graphite sample was oxidised. The maximum value of the release rate ratio was obtained for tritium in experiments with argon atmosphere at 1060°C. It was shown that 14C could be separated from the main graphite mass with a 14C/12C enrichment factor of about 20. However, the total fractional release was not sufficient for a pilot scale process under these conditions. The fractional release of 14C can be increased by addition of water steam, but in this case the enrichment factor drops below 5. In general, this separation of carbon isotopes is only possible because 14C is mainly located near the grain boundary surfaces and its concentration profile decreases with depth inside the graphite grains. The radionuclides 60Co, 154,155Eu, 134,137Cs and 133Ba present in contaminated graphite mainly remained in the ceramic reaction boat in contrast to the volatile Cs. By optimising the process parameters this decontamination process may be developed further into a pilot scale technology for graphite purification with fractional reduction of the 14C inventory.","abstract_html":"The present work deals with radioactive graphite management. Two different aspects were examined: 1. The corrosion behaviour of graphite under final repository conditions 2. The decontamination of graphite by thermal treatment Spent fuel from the German AVR and THTR high-temperature reactors is envisaged for direct disposal in a final repository. In this respect, corrosion of the graphite matrix of the spherical fuel elements and pyrocarbon coatings from coated fuel particles is a very important aspect for predicting the long-term behaviour of the radioactive waste. The graphite materials to be examined were isolated from unirradiated fuel elements, and their corrosion was investigated in different aqueous phases. The experiments were performed in deionised water, MgCl2-rich (brine-2) and NaCl-rich (brine-3) solutions at 90°C under argon, oxygen and air atmosphere. In order to investigate the influence of aqueous phase radiolysis, further experiments were performed in the presence of g-irradiation sources under argon atmosphere. It was established that the slow corrosion of carbonaceous materials in different aquatic phases in the absence of irradiation was caused by interaction with dissolved oxygen. Corrosion rates in aqueous solutions under pure oxygen and air atmosphere decrease in the order water > brine?3 > brine-2. Acceleration of graphite and pyrocarbon corrosion was observed in the presence of g-irradiation in brines while the opposite influence of the aqueous phases on the corrosion rates was additionally observed. The corrosion rates of graphite and pyrocarbon in irradiated brine-2 and brine-3 in argon atmosphere were two orders of magnitude higher than in pure oxygen atmosphere without irradiation. This may be related to the formation of highly oxidising chlorine species in brines, which react with carbon materials. In pure water, radiolysis did not influence the oxidation process significantly. However, calculating the expected lifetime of the graphite matrix in the repository on the basis of the corrosion rates determined is an extremely conservative consideration since the radiation dose rates used are significantly higher than could ever occur during final disposal of high-temperature reactor fuel elements. In the second part of the present work, thermal treatment of contaminated structural graphite from the high-temperature reactor core and thermal columns of the research reactors was investigated as a possible decontamination process. The main problem associated with direct disposal of contaminated graphite is its large volume. Reprocessing of the graphite based on graphite gasification offers the opportunity to separate the radionuclides from the main graphite mass, which could then be reused or disposed conventionally. The experiments were performed in an argon flow and steam in the temperature range of 870 – 1060°C. Comparison of the release rate ratios of the volatilised radionuclides 14C and 3H with the release of 12C showed that under all experimental conditions tritium and 14C were released faster than the graphite sample was oxidised. The maximum value of the release rate ratio was obtained for tritium in experiments with argon atmosphere at 1060°C. It was shown that 14C could be separated from the main graphite mass with a 14C/12C enrichment factor of about 20. However, the total fractional release was not sufficient for a pilot scale process under these conditions. The fractional release of 14C can be increased by addition of water steam, but in this case the enrichment factor drops below 5. In general, this separation of carbon isotopes is only possible because 14C is mainly located near the grain boundary surfaces and its concentration profile decreases with depth inside the graphite grains. The radionuclides 60Co, 154,155Eu, 134,137Cs and 133Ba present in contaminated graphite mainly remained in the ceramic reaction boat in contrast to the volatile Cs. By optimising the process parameters this decontamination process may be developed further into a pilot scale technology for graphite purification with fractional reduction of the 14C inventory.","abstract_has_math":false,"creators":["Podruzhina, Tatiana"],"institution":"Publikationsserver der RWTH Aachen University","degree_name":null,"degree_level":null,"degree_discipline":null,"degree_department":null,"school":null,"contributors":["Odoj, Reinhard"],"advisors":[],"committee_chairs":[],"committee_members":[],"year":2004,"date_issued":"2004","date_published":"2004","updated_at":"2026-07-30T19:40:42Z","subjects":["info:eu-repo/classification/ddc/620","Graphit","Radioaktiver Abfall","Endlagerung","Elektrochemische Korrosion","Korrosionsprüfung","Dekontamination","Wärmebehandlung","Ingenieurwissenschaften","graphite","radioactive waste","corrosion","thermal treatment"],"languages":["eng"],"rights":["info:eu-repo/semantics/openAccess"],"rights_urls":[],"identifier_entries":[{"key":"dc:identifier","label":"Identifier","values":["https://publications.rwth-aachen.de/search?p=id:%22RWTH-CONV-114122%22"],"render_values":[{"text":"https://publications.rwth-aachen.de/search?p=id:%22RWTH-CONV-114122%22","href":"https://publications.rwth-aachen.de/search?p=id:%22RWTH-CONV-114122%22","code":true}]}]},"links":{"outbound_url":"https://publications.rwth-aachen.de/record/51873","outbound_label":"Repository record","outbound_source":"dc:identifier"},"source_record":{"url":"https://publications.rwth-aachen.de/oai2d?verb=GetRecord&metadataPrefix=oai_dc&identifier=oai%3Apublications.rwth-aachen.de%3A51873","prefix":"oai_dc"},"metadata_groups":[{"id":"people","label":"People","entries":[{"key":"dc:contributor","label":"Contributor","values":["Odoj, Reinhard"]},{"key":"dc:creator","label":"Author","values":["Podruzhina, Tatiana"]}]},{"id":"academic_context","label":"Academic Context","entries":[{"key":"dc:coverage","label":"Dc Coverage","values":["DE"]},{"key":"dc:date","label":"Dc Date","values":["2004"]},{"key":"dc:publisher","label":"Institution","values":["Publikationsserver der RWTH Aachen University"]},{"key":"dc:relation","label":"Dc Relation","values":["info:eu-repo/semantics/altIdentifier/urn/urn:nbn:de:hbz:82-20050346"]},{"key":"dc:type","label":"Dc Type","values":["info:eu-repo/semantics/doctoralThesis","info:eu-repo/semantics/publishedVersion"]}]},{"id":"subjects_keywords","label":"Subjects and Keywords","entries":[{"key":"dc:subject","label":"Dc Subject","values":["info:eu-repo/classification/ddc/620","Graphit","Radioaktiver Abfall","Endlagerung","Elektrochemische Korrosion","Korrosionsprüfung","Dekontamination","Wärmebehandlung","Ingenieurwissenschaften","graphite","radioactive waste","corrosion","thermal treatment"]}]},{"id":"language_rights","label":"Language and Rights","entries":[{"key":"dc:language","label":"Dc Language","values":["eng"]},{"key":"dc:rights","label":"Dc Rights","values":["info:eu-repo/semantics/openAccess"]}]},{"id":"identifiers","label":"Identifiers","entries":[{"key":"dc:identifier","label":"Identifier","values":["https://publications.rwth-aachen.de/record/51873","https://publications.rwth-aachen.de/search?p=id:%22RWTH-CONV-114122%22"]}]},{"id":"additional","label":"Additional Metadata","entries":[{"key":"dc:description","label":"Description","values":["The present work deals with radioactive graphite management. Two different aspects were examined: 1. The corrosion behaviour of graphite under final repository conditions 2. The decontamination of graphite by thermal treatment Spent fuel from the German AVR and THTR high-temperature reactors is envisaged for direct disposal in a final repository. In this respect, corrosion of the graphite matrix of the spherical fuel elements and pyrocarbon coatings from coated fuel particles is a very important aspect for predicting the long-term behaviour of the radioactive waste. The graphite materials to be examined were isolated from unirradiated fuel elements, and their corrosion was investigated in different aqueous phases. The experiments were performed in deionised water, MgCl2-rich (brine-2) and NaCl-rich (brine-3) solutions at 90°C under argon, oxygen and air atmosphere. In order to investigate the influence of aqueous phase radiolysis, further experiments were performed in the presence of g-irradiation sources under argon atmosphere. It was established that the slow corrosion of carbonaceous materials in different aquatic phases in the absence of irradiation was caused by interaction with dissolved oxygen. Corrosion rates in aqueous solutions under pure oxygen and air atmosphere decrease in the order water > brine?3 > brine-2. Acceleration of graphite and pyrocarbon corrosion was observed in the presence of g-irradiation in brines while the opposite influence of the aqueous phases on the corrosion rates was additionally observed. The corrosion rates of graphite and pyrocarbon in irradiated brine-2 and brine-3 in argon atmosphere were two orders of magnitude higher than in pure oxygen atmosphere without irradiation. This may be related to the formation of highly oxidising chlorine species in brines, which react with carbon materials. In pure water, radiolysis did not influence the oxidation process significantly. However, calculating the expected lifetime of the graphite matrix in the repository on the basis of the corrosion rates determined is an extremely conservative consideration since the radiation dose rates used are significantly higher than could ever occur during final disposal of high-temperature reactor fuel elements. In the second part of the present work, thermal treatment of contaminated structural graphite from the high-temperature reactor core and thermal columns of the research reactors was investigated as a possible decontamination process. The main problem associated with direct disposal of contaminated graphite is its large volume. Reprocessing of the graphite based on graphite gasification offers the opportunity to separate the radionuclides from the main graphite mass, which could then be reused or disposed conventionally. The experiments were performed in an argon flow and steam in the temperature range of 870 – 1060°C. Comparison of the release rate ratios of the volatilised radionuclides 14C and 3H with the release of 12C showed that under all experimental conditions tritium and 14C were released faster than the graphite sample was oxidised. The maximum value of the release rate ratio was obtained for tritium in experiments with argon atmosphere at 1060°C. It was shown that 14C could be separated from the main graphite mass with a 14C/12C enrichment factor of about 20. However, the total fractional release was not sufficient for a pilot scale process under these conditions. The fractional release of 14C can be increased by addition of water steam, but in this case the enrichment factor drops below 5. In general, this separation of carbon isotopes is only possible because 14C is mainly located near the grain boundary surfaces and its concentration profile decreases with depth inside the graphite grains. The radionuclides 60Co, 154,155Eu, 134,137Cs and 133Ba present in contaminated graphite mainly remained in the ceramic reaction boat in contrast to the volatile Cs. By optimising the process parameters this decontamination process may be developed further into a pilot scale technology for graphite purification with fractional reduction of the 14C inventory."]},{"key":"dc:source","label":"Dc Source","values":["Aachen : Publikationsserver der RWTH Aachen University II, 115 S. : Ill., graph. Darst. (2004). = Aachen, Techn. Hochsch., Diss., 2004"]},{"key":"dc:title","label":"Title","values":["Graphite as radioactive waste : corrosion behaviour under final repository conditions and thermal treatment"]}]}],"canonical_facts":{"dc:contributor":["Odoj, Reinhard"],"dc:coverage":["DE"],"dc:creator":["Podruzhina, Tatiana"],"dc:date":["2004"],"dc:description":["The present work deals with radioactive graphite management. Two different aspects were examined: 1. The corrosion behaviour of graphite under final repository conditions 2. The decontamination of graphite by thermal treatment Spent fuel from the German AVR and THTR high-temperature reactors is envisaged for direct disposal in a final repository. In this respect, corrosion of the graphite matrix of the spherical fuel elements and pyrocarbon coatings from coated fuel particles is a very important aspect for predicting the long-term behaviour of the radioactive waste. The graphite materials to be examined were isolated from unirradiated fuel elements, and their corrosion was investigated in different aqueous phases. The experiments were performed in deionised water, MgCl2-rich (brine-2) and NaCl-rich (brine-3) solutions at 90°C under argon, oxygen and air atmosphere. In order to investigate the influence of aqueous phase radiolysis, further experiments were performed in the presence of g-irradiation sources under argon atmosphere. It was established that the slow corrosion of carbonaceous materials in different aquatic phases in the absence of irradiation was caused by interaction with dissolved oxygen. Corrosion rates in aqueous solutions under pure oxygen and air atmosphere decrease in the order water > brine?3 > brine-2. Acceleration of graphite and pyrocarbon corrosion was observed in the presence of g-irradiation in brines while the opposite influence of the aqueous phases on the corrosion rates was additionally observed. The corrosion rates of graphite and pyrocarbon in irradiated brine-2 and brine-3 in argon atmosphere were two orders of magnitude higher than in pure oxygen atmosphere without irradiation. This may be related to the formation of highly oxidising chlorine species in brines, which react with carbon materials. In pure water, radiolysis did not influence the oxidation process significantly. However, calculating the expected lifetime of the graphite matrix in the repository on the basis of the corrosion rates determined is an extremely conservative consideration since the radiation dose rates used are significantly higher than could ever occur during final disposal of high-temperature reactor fuel elements. In the second part of the present work, thermal treatment of contaminated structural graphite from the high-temperature reactor core and thermal columns of the research reactors was investigated as a possible decontamination process. The main problem associated with direct disposal of contaminated graphite is its large volume. Reprocessing of the graphite based on graphite gasification offers the opportunity to separate the radionuclides from the main graphite mass, which could then be reused or disposed conventionally. The experiments were performed in an argon flow and steam in the temperature range of 870 – 1060°C. Comparison of the release rate ratios of the volatilised radionuclides 14C and 3H with the release of 12C showed that under all experimental conditions tritium and 14C were released faster than the graphite sample was oxidised. The maximum value of the release rate ratio was obtained for tritium in experiments with argon atmosphere at 1060°C. It was shown that 14C could be separated from the main graphite mass with a 14C/12C enrichment factor of about 20. However, the total fractional release was not sufficient for a pilot scale process under these conditions. The fractional release of 14C can be increased by addition of water steam, but in this case the enrichment factor drops below 5. In general, this separation of carbon isotopes is only possible because 14C is mainly located near the grain boundary surfaces and its concentration profile decreases with depth inside the graphite grains. The radionuclides 60Co, 154,155Eu, 134,137Cs and 133Ba present in contaminated graphite mainly remained in the ceramic reaction boat in contrast to the volatile Cs. By optimising the process parameters this decontamination process may be developed further into a pilot scale technology for graphite purification with fractional reduction of the 14C inventory."],"dc:identifier":["https://publications.rwth-aachen.de/record/51873","https://publications.rwth-aachen.de/search?p=id:%22RWTH-CONV-114122%22"],"dc:language":["eng"],"dc:publisher":["Publikationsserver der RWTH Aachen University"],"dc:relation":["info:eu-repo/semantics/altIdentifier/urn/urn:nbn:de:hbz:82-20050346"],"dc:rights":["info:eu-repo/semantics/openAccess"],"dc:source":["Aachen : Publikationsserver der RWTH Aachen University II, 115 S. : Ill., graph. Darst. (2004). = Aachen, Techn. Hochsch., Diss., 2004"],"dc:subject":["info:eu-repo/classification/ddc/620","Graphit","Radioaktiver Abfall","Endlagerung","Elektrochemische Korrosion","Korrosionsprüfung","Dekontamination","Wärmebehandlung","Ingenieurwissenschaften","graphite","radioactive waste","corrosion","thermal treatment"],"dc:title":["Graphite as radioactive waste : corrosion behaviour under final repository conditions and thermal treatment"],"dc:type":["info:eu-repo/semantics/doctoralThesis","info:eu-repo/semantics/publishedVersion"]},"updated_at":"2026-07-30T19:40:42Z"}