GERMAN PLANNINGS ON CO-LOCATION OF LLW, ILW,
HLW AND SPENT FUEL
Peter W. Brennecke/Berndt-Rainer Martens/Günther W. Tittel
Bundesamt für Strahlenschutz, Salzgitter, Germany
(Federal Office for Radiation Protection)
ABSTRACT
Since the early sixties, the radioactive waste disposal policy in the Federal Republic of Germany has been based on the decision that all types of radioactive wastes are to be disposed of in deep geological formations. Near-surface disposal is not practised in Germany because of the high population density, climatic conditions and the existence of appropriate deep geological formations.
The Gorleben salt dome in the northern part of Germany is being investigated for its suitability to host a repository at depths between 840 m and 1,200 m for all types of radioactive waste, mainly for high-level and/or alpha-bearing waste from reprocessing and spent fuel elements. Site investigations and planning work are at present concentrated in a first step on the north-eastern part of the salt dome in order to find out this part being sufficient for the disposal of radioactive wastes originating from nuclear energy use in Germany. The accumulated inventory of beta/gamma and alpha emitters to be emplaced within an operational period of about 70 years is estimated to be in the order of magnitude of 1021 Bq and 1019 Bq, respectively.
INTRODUCTION
The objective of disposal in repositories is to ensure that radioactive waste is handled and stored in such a way that the protection of man and the environment from harm caused by the ionizing radiation of the waste is guaranteed. Consequently, the planning and construction of a repository must be carried out in such a way that this objective will be met during its operational and post-closure phase. Specific challenges will occur provided that e. g. high-level waste (HLW) and/or alpha-bearing waste originating from reprocessing and spent fuel are intended to be disposed of. Thus, the very long half-lives and other specific disposal-related properties of these types of radioactive waste must be taken into account when demonstrating the observance of the objective of disposal.
In the Federal Republic of Germany, the disposal of radioactive waste is planned or has been carried out in deep geological formations only. In July 1977, the Gorleben salt dome situated in the northern part of Germany was designated as a site for the construction and operation of a repository. It is intended to dispose of all types of radioactive waste in this facility, in particular heat-generating waste (i.e., HLW) from reprocessing and spent fuel. The selection of salt as host rock for the disposal of HLW is in accordance with appropriate international findings. According to this intention LLW, ILW, HLW and spent fuel are planned to be disposed of in one repository.
GERMAN RADIOACTIVE WASTE DISPOSAL POLICY
In the Federal Republic of Germany, the use of nuclear energy especially started with the operation of the first nuclear power plant in 1960. Since the early sixties, i. e. from its very beginning, the German radioactive waste disposal policy has been based on the decision that all types of radioactive waste are to be disposed of in deep geological formations. Such a decision is only realistically acceptable if a barrier to radionuclide releases exists which remains effective over very long periods of time, which is needed by radionuclides to decay significantly. Thus, vitrified fission product solution originating from reprocessing and spent fuel originating from power reactors, research reactors or material test reactors as well as spent sealed radiation sources and miscellaneous waste from small waste generators are affected by this decision. It also applies to alpha-bearing waste originating in particular from reprocessing facilities, nuclear research establishments or the nuclear fuel cycle industry. Near-surface disposal or shallow land burial is not practised in Germany because of high population density, climatic conditions and existing appropriate deep geological formations.
LEGAL PRINCIPLES
The disposal of radioactive waste in a repository is in particular governed by the following specific acts and regulations:
The protection objectives of radioactive waste disposal in a repository are prescribed by the Atomic Energy Act and the Radiation Protection Ordinance. The Federal Mining Act regulates all aspects concerning the operation of a disposal mine. The Safety Criteria specify the measures to be taken in order to achieve that this objective has been reached.
The peaceful use of nuclear energy in Germany is regulated by the Atomic Energy Act. On September 5, 1976, its Fourth Amendment was enacted. It provided the legal basis for the disposal of radioactive waste. According to section 9 a of this act, the Federal Government has to establish installations for the engineered storage and disposal of radioactive waste, i. e. the disposal of radioactive waste is assigned to the Federal Government as a sovereign task. On November 1, 1989, this competence was assigned to the Bundesamt für Strahlenschutz (BfS, Federal Office for Radiation Protection). Accordingly, BfS is responsible for the establishment and operation of those federal installations, acting on behalf of the Federal Government.
The legal competences for the licensing of the construction and operation of a repository are regulated in such a way that two procedures must be performed: on the one hand, the procedure under atomic law and, on the other hand, the procedure under mining law. For the establishment of a repository, pursuant to section 9b of the Atomic Energy Act, the initiation of a plan-approval procedure, i. e. a special kind of a licensing procedure, has to be applied to the respective licensing authority of the federal state [1]. BfS is the authorized applicant.
SAFETY CRITERIA
The basic aspects which must be taken into account to achieve the objective of disposal are compiled in the "Safety Criteria for the Disposal of Radioactive Waste in a Mine" [2]. Their scope implies all types of radioactive waste to be disposed of. The safety criteria qualitatively specify the measures to be taken in order to achieve the protection objective of disposal and define the principles by which it must be demonstrated that this objective has been reached, i. e. technical measures and methods of procedure are to be adjusted to one another. The importance of the site selection, the system consisting of geology/repository/waste packages, the multibarrier concept, and the use of state-of-the-art technology are emphasized. The following criteria are considered to be the most important ones:
The safety criteria permit a certain latitude of judgement. Such margins gradually diminish in the realization of a repository project. This process is predominantly determined by a site-specific safety assessment, within the scope of which the required safety of the repository must quantitatively be demonstrated including the derivation of requirements on the design of the facility as well as on the waste packages to be disposed of.
Nevertheless, the protection objective can only be achieved by an iterative process drawing together more and more detailed information obtained as the respective repository project progresses through its various phases of investigation, planning, detailed design and performance assessment, thus becoming more and more concrete.
The safety criteria comprise the most important features characterizing the German approach to disposal (basic concept) and the respective philosophy employed:
The Safety Criteria for the Disposal of Radioactive Wastes in a Mine issued in 1983 [2] are at present revised on behalf of the Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit (BMU, Federal Ministry for the Environment, Nature Conservation and Nuclear Safety) being the competent federal authority for nuclear safety, radiation protection and waste management in Germany.
RADIOACTIVE WASTE ORIGINS AND ARISINGS
In Germany, 19 nuclear power plants with a gross generating capacity of about 22 GW are presently in operation. 13 of them are PWRs, six are BWRs. Reprocessing of spent fuel elements originating from these plants takes place abroad in French and British facilities. Basic and applied investigations are performed in several nuclear research establishments, uranium enrichment and fuel element fabrication facilities are operated, and nuclear facilities are decommissioned and dismantled. Finally, there are many smaller waste generators, e. g. universities, industrial companies, hospitals, medical centres, the German Federal Armed Forces, and pharmaceutical and biomedical companies.
As to the waste arisings, BfS carries out an annual inquiry into the amounts of unconditioned and conditioned waste in Germany. According to the 1995 inquiry [3], the amounts of untreated/unconditioned LLW/ILW and HLW were about 30,100 m3 and 500 m3, respectively. The existing amounts of conditioned LLW/ILW and HLW (without spent fuel) totalled to about 63,000 m3 and 1,900 m3, respectively, on December 31, 1995. The annual arisings of conditioned LLW/ILW amounts on average to 4,300 m3. Besides, a forecast into future waste arisings until 2080 resulted in about 412,400 m3 for conditioned LLW/ILW (of this, about 40,000 m3 will be disposed of in the Morsleben repository) and about 51,300 m3 HLW including spent fuel.
THE GORLEBEN REPOSITORY PROJECT
According to the German radioactive waste disposal planning work [4], the Gorleben salt dome located in the northern part of Germany is at present investigated for its suitability to host a repository for all types of solid and solidified radioactive waste including especially the emplacement of heat-generating waste, i. e. HLW such as vitrified fission product solution or high-force compacted hulls and ends originating from reprocessing. Beyond this, the direct disposal of spent fuel elements is included according to the amendment of the Atomic Energy Act in July 1994 providing the legal basis for direct disposal of spent fuel as alternative to reprocessing. Thus, co-location of LLW, ILW, HLW and spent fuel is planned in this geological repository.
In July 1977, the State Government of Lower Saxony designated Gorleben as a provisional site of the formerly waste management centre including a disposal mine. The selection of this site for the construction and operation of a repository was especially based on screening investigations using the knowledge on salt domes at that time.
As the underground exploration of the Gorleben salt dome is still in its beginnings, the existing conceptual design of the repository depends on model assumptions [5]. The planning of the underground facilities was first based on a site-independent generic model of a repository within a salt dome, reflecting the known size, shape and depth of the Gorleben salt dome only. The conceptual design was replaced by module-like design alternatives, taking into account the present site-specific knowledge, different emplacement methods, and actualized data on the radioactive waste intended for disposal. Thus, it will be necessary to adapt the final layout of this facility at the end of the underground investigations.
According to the present model layout of the Gorleben repository, sufficient disposal capacity will be available to guarantee the disposal of all types of solid and solidified radioactive waste originating from nuclear energy use in Germany. Only radioactive waste appropriately conditioned, i.e. waste packages, will be accepted for disposal; gaseous and liquid waste is not considered suitable for emplacement in a geological repository.
The operational lifetime of this disposal facility is estimated to be about 70 years. The accumulated inventory of beta/gamma emitters and alpha emitters to be disposed of during this period of time is estimated to be in the order of magnitude of 1021 Bq and 1019 Bq, respectively.
At present, a disposal mine at an emplacement level depth of about 870 m has been planned. According to the safety criteria [2], the number of shafts is to be kept at a minimum, although at least two shafts are required for transport, ventilation and safety reasons. The shafts are located in the centre of the underground facilities still to be mined. Their distance is about 500 m; no waste packages will be disposed of within a safety pillar of 300 m around them. The underground exploration of the salt dome's interior will be performed about 30 m above the later emplacement level at 870 m; later on the exploration drifts are intended to be used for ventilation purposes (return air). The excavation of drifts, galleries and disposal rooms is to be performed with techniques identical to those successfully applied in salt mining operations. According to the safety criteria, only well-tried and proven mining methods as well as state-of-the-art technology are to be used.
Radioactive waste with negligible heat generation, i.e. LLW and ILW, is planned to be emplaced in disposal rooms with a cross-section of about 40 m2 and a length of about 200 m using the stacking technique. The heat-generating radioactive waste, i. e. vitrified fission product solution in canisters, is planned to be disposed of in vertical boreholes up to 300 m deep. This emplacement technique makes use of the probably more vertical extension of the Halite body in the salt dome and the fact that there will be a short-term convergence of the small annular gap in filled boreholes due to the heat generation of the respective waste packages. In addition to that, emplacement into inclined boreholes about 30 m deep is under consideration. Furthermore, besides handling problems, the emplacement of heat-generating waste packages such as canisters in disposal rooms or drifts would probably cause an inadmissible thermal impact upon the host rock. It should be added that the thermal layout of the repository is based on an assumed maximum temperature of 200° C at the surface of the canisters. This limitation depends upon, among other things, the age of the waste, in particular geomechanical aspects, the distance and length of the boreholes and the emplacement of waste packages within the boreholes (longitudinal heat dilution). In the case of direct disposal of spent fuel, two possibilities are discussed at present. On the one hand, spent fuel in thick-walled self-shielding casks are intended to be emplaced in galleries. On the other hand, spent fuel elements in specially designed thin-walled unshielded packagings are intended to be emplaced in vertical or inclined boreholes, too.
In consideration of safety-related requirements, the emplacement of waste packages will be done in retreat work from the boundaries of the disposal fields in direction to the shafts. Backfilling, plugging and sealing of disposal rooms, boreholes and disposal fields as well as of both shafts is considered one of the most important engineered safety measures.
SITE INVESTIGATIONS
Pursuant to the German safety criteria for the disposal of radioactive waste and due to common practice, a detailed site characterization programme including above-ground and underground investigations must be performed. The Gorleben site characterization programme started in 1979. The investigations include the determination of lithological, facial and stratigraphical properties as well as the structural and hydrological situation of the overlying strata of the salt dome. The contact zone between salt dome and overlying strata was particularly explored.
The following work from above ground was carried out between 1979 and 1985 to investigate the geology and hydrogeology of the Gorleben site:
Other investigations which were carried out include geoelectrical and geothermal studies, gravimetry, seismology, geochemistry, isotope geochemistry, and micropalaeontology. The investigation of the site from above ground will be continued for some time. To replace assumptions and preliminary data in the hydrogeological modelling of the site, territory of the former German Democratic Republic (65 km2 out of a total of 350 km2) north of the river Elbe is now included in the site investigation.
The underground investigation programme started in 1984. The objective of the exploration from underground is to acquire all information needed to assess the operational and long-term safety of the planned repository. This has to be achieved while keeping disadvantages of potential damage to the geological barrier as low as reasonably achievable. The sinking of two shafts began in 1986. After finishing the drift connecting both shafts in October 1996, excavation of the exploratory mine was started at a depth of 840 m with the construction of operational rooms. The exploratory shafts were sunk through the Cenozoic cover and the cap rock into the rock salt up to their final depth of 933 m and 843 m, respectively. Subsequently two pairs of exploratory drifts, connected by eight cross-cuts are presently being driven to the northeast in a depth of 840 m. From there, numerous exploratory wells (in total about 60 km) will be drilled horizontally and vertically up to about the outer boundary of the prospective repository fields.
The stratigraphy and structure of the salt dome will be investigated by geological, geophysical and petrographic methods in such detail that it will be possible to identify rock salt sufficiently large and otherwise suitable for the different types of radioactive waste. Also, the positions of the more problematic layers, such as the main anhydrite and the Stassfurt potash seam, as well as brine pockets and gas-bearing salt bodies, will have to be determined exactly for proper assessments of the Gorleben salt dome´s suitability.
It is expected that the characterization and subsequent assessment of the Gorleben site will take several years and, according to present plannings, will be finished some years after the turn of the century. Up to now, the results of the surface exploration programme and the first results of the geoscientific underground investigation programme substantiate the potential suitability of the Gorleben site, i. e. the expectation of being able to reach the objectives of disposal by long-term retention and dilution and, thus, to host a repository for HLW, alpha-bearing waste and spent fuel. A final judgement, however, has to be based mainly on the final results from the underground investigations.
SITE-SPECIFIC SAFETY ASSESSMENTS
Pursuant to the safety criteria [2], the safety of the planned Gorleben repository must be proved within the scope of a site-specific safety assessment. On the basis of such an assessment covering the total geological and hydrogeological situation, the technical design of this facility including its anticipated mode of operation and the waste packages intended to be disposed of, the safety of the Gorleben repository during the operational and post-closure phase has to be demonstrated. The site-specific safety assessment to be carried out comprises the normal operation of the planned facility, assumed incidents, the thermal influence upon the host rock, the nuclear criticality safety and the radiological long-term safety in the post-closure phase. Of this, nuclear criticality safety, long-term safety and human intrusion shall be addressed in more detail.
Criticality Safety Considerations
General criticality with regard to the direct disposal of spent fuel elements was examined several years ago. It has been shown that, from the viewpoint of criticality safety, direct disposal is a possible option. Another option which is part of the disposal strategy for the planned Gorleben repository, is the emplacement of rods (i.e. disassembled fuel bundles) packaged in canisters. For preliminary criticality considerations it is assumed that the rods remain structurally intact such that the enclosure of the radioactive material is ensured. A tight package of fuel rods influences the essential parameters for criticality safety. The criticality considerations were restricted to the determination of the infinite multiplication factor; water access to the fuel rods was assumed, conservatively neglecting the fact that the realistic moderator in the repository is a salt solution or brine. As a result it has been calculated that the fuel elements are more reactive than closely packed bundles of fuel rods owing to the near optimal moderator to fuel ratio. Hence, the disassembling of the fuel elements and the tight package of the rods may be favourable in respect of criticality safety.
Long-term Safety Aspects
Because of its favourable properties rock salt is the main long-term barrier of the planned repository in the Gorleben salt dome. Nevertheless, the possible release of radionuclides during the post-closure phase has been investigated in a conservative approach assuming that brines may intrude into the backfilled parts of the repository. Both the creation of pathways for water in anhydrite horizons due to thermomechanical effects caused by the heat-generating waste and the existence of brine inclusions in the rock salt were considered.
Possible releases of radionuclides via the water path are assessed and the respective dose rates calculated. Comprehensive experimental investigations on site-specific samples provide the necessary database for the performance of a quantitative long-term safety assessment for the Gorleben repository project. The sorption and desorption experiments have been focused on the influence of natural groundwater colloids on the sorption and transport behaviour of radionuclides, especially on elements such as Np, Pu, Am and Cm. The latest experiments supplement previous investigations, aiming to deepen the understanding of the sorption behaviour and the underlying sorption mechanisms [6]. The results in general confirm the sorption data of radionuclides and their dependence on the influencing parameters as, e. g., Eh, ph, natural and artificial complexing agents as humic substances and EDTA and colloid formation.
With the completion of the experiments on migration, characteristic sorption data for the relevant elements are now available for 56 sediment and ground water systems. The quantification of chemical variables which affect sorption was made possible by means of a targeted variation of the input parameters. The investigations into the radionuclide sorption for Np, Pu and Am have shown that sorption is generally higher in cohesive sediments compared to sandy sediments, i.e., in the range of orders of magnitude. An influence of humic substances on the sorption characteristics of radionuclides is identifiable, particularly in the sandy systems. Where the content in humic substances increases, sorption by sediments decreases. The redox potential of the sediment and ground water systems constitutes, for Np sorption, the basis of the most significant parameters. Where reductive conditions were present and redox potential was low, a higher level of sorption by the sediment was measured for Np.
In order to complete the migration experiments, microbiological investigations were carried out, in particular aiming at
Since microorganisms were found to be present in very low numbers, these presumably play only a secondary role within the dispersion of radionuclides. Beyond this, due to the non-sterile conditions of the sorption experiments, an influence due to microorganisms was detected.
The results of these investigations contribute to the data base for model calculations still to be performed, aiming at the demonstration of a long-term safety of the Gorleben site.
Human Intrusion
Within the scope of the Gorleben site-specific safety assessment still to be performed, human intrusion will be addressed, too. The safety criteria do not deal with unintentional human intrusion. It is basically assumed that a long-term information preservation given by the documentation on the repository and the waste packages disposed of, in connection with appropriate administrative measures, will contribute to prevent unintentional human intrusion. Nevertheless, as a future loss of such information and data cannot be excluded, this intrusion, e. g. by drilling, will have to be taken into account.
Within the licensing procedure for the Konrad repository project, unintentional human intrusion by drilling was considered, taking site and facility-specific characteristics into account. It could be shown that the resulting radiation exposure would to be expected in the order of magnitude of the exposure permissible to radiation exposed personnel (cf. section 49 of the Radiation Protection Ordinance). Thus, unintentional human intrusion of the Konrad repository was assigned to the residual risk which is to be neglected
RADIATION PROTECTION DURING THE POST-CLOSURE PHASE
The radiation protection objectives for a repository are prescribed and quantified in the Strahlenschutzverordnung (Radiation Protection Ordinance). According to the revised ordinance of 1989,
Thus, regulatory dose limits have been set which must be complied with. Evidence of this protection objective, according to the safety criteria [2], must be demonstrated within the radiological long-term safety assessment. By this means, possible radiation exposures to individuals will be kept within the variability of natural radiation rates. Nevertheless, such a procedure is only reasonable for periods of time for which changes in the geological barriers and in man's environment can still be forecast with sufficient reliability.
According to the common statement of the Reaktor-Sicherheitskommission (Reactor Safety Commission) and the Strahlenschutzkommission (Commission on Radiological Protection), given on behalf of BMU within the scope of the licensing procedure for the Konrad repository, the radiological long-term safety assessment covering the calculation of individual doses is limited to time periods in the order of 10,000 years. Beyond this time period, i. e. up to 1,000,000 years, the isolation potential of the geological system of the chosen site may be assessed. This allows adequate safety margins. Nevertheless, this statement was not accepted by the competent licensing authority for the Konrad repository project and, as a result, radiation exposure calculations were carried out covering a period of time of more than 10,000,000 years. It is to be expected that the same time period will have to be applied within the Gorleben radiological long-term safety assessment.
To ensure the long-term safety of the planned Gorleben repository including the proof that, i. e. the individual dose limits of 0.3 mSv/a or 0.9 mSv/a are complied with, any possible release of radionuclides via water path must be assessed and the respective dose rates be calculated. For this purpose, it is necessary to determine all the relevant basic data required for the performance of the long-term safety assessment. This includes, among other things, the results of the underground investigation of this site, the termination of which is currently expected by the year 2005, together with an assessment of its suitability to host a repository for the disposal of all types of radioactive waste.
DECOMMISSIONING AND POST-CLOSURE
The planned Gorleben repository will be decommissioned after the operational phase. Parts of the underground facility, e. g. disposal rooms and boreholes filled with waste packages, will already be shut down during the operational phase. The decommissioning will be completed with the filling and sealing of both shafts. Filling and closing off the mine openings have as a goal the increased stability by means of reduction of remaining voids, thereby delaying or hindering the access of transport media (e. g., water or brines) to the radioactive wastes, and minimizing a radionuclide release to a permissible level. As to the safety criteria [2], these measures are the final contribution to the long-term safety of this repository.
The safety criteria require that construction, operation, and decommissioning of the repository are to be performed and monitored such that no particular control or monitoring programme is necessary during the post-closure phase. Routinely performed, general environmental measurements as well as topographic measurements will give information on the radiology and the long-term thermo-mechanical behaviour of, e.g., the host rock and the overburden.
Data on the repository, the waste packages disposed of, and the essential technical measures taken during construction, operation and decommissioning should be documented. Complete documentation should be maintained at suitable separate locations. A surface marker for the planned Gorleben repository is not necessary taking the normal environmental protection and topographical measurements into consideration. Knowledge of the site's location should be guaranteed sufficiently by the documentation.
REALIZATION OF THE GORLEBEN REPOSITORY PROJECT
In order to demonstrate the required safety of the Gorleben repository during the operational and the post-closure phase, a stepwise procedure was agreed upon. In the first step, using assumptions and model data on the Gorleben salt dome, the disposability of all types of radioactive waste, i.e. LLW, ILW and HLW and spent fuel, was basically be proven. Essential prerequisites for the second step are the final confirmation of the salt dome´s suitability, detailed plannings of the above-ground and underground facilities and an updating of the radioactive waste data base.
REFERENCES