E. DEEP GEOLOGICAL REPOSITORY SAFETY

1. How is the safety of the deep geological repository ensured, or leakage of radioactive substances into the environment prevented?

The safety of the DGR after its closure is ensured by a multi-barrier approach to the isolation of the disposed RAW, i.e. by the use of a series of redundant, independent safety barriers that prevent the penetration of radioactivity into environmental compartments.

In the Czech disposal policy, the main barriers are a two-layered disposal packaging surrounded by compacted bentonite and, to the maximum extent possible, undisturbed rock at a depth of approximately 500 m below the surface. A system of interconnected barriers prevents radionuclides from entering the environment. The safety of the disposal facility is planned in such a way that even if one of the barriers fails for any reason, the safety of the disposal facility cannot be compromised, i.e. the other barriers prevent the release of radioactive substances into the environment.

The safety of the DGR is also ensured by safety during the operation and closure of the DGR. The safety assessment generally assesses the level of nuclear safety, radiation protection, technical safety, radiation monitoring, radiation extraordinary event management and security in accordance with the provisions of Section 48 of Act No. 263/2016 Coll. and Decree No. 162/2017 Coll.

2. Is it even possible to predict the long-term safety of the DGR? Who verifies the safety of the disposal facility and who ensures it?

Yes, it is possible to predict the long-term safety of DGR using mathematical models used in computer programs. The mathematical modelling is based on the basic laws of physics and the calculation programmes must be verified and validated in accordance with the requirement of Article 9(2)(b) of Decree No 377/2016 Coll.; in the case of validation, mainly by laboratory and in-situ experiments and by using knowledge from so-called natural analogues. The safety of each NPP, including the DGR, is the responsibility of its operator, i.e. in the case of the DGR it will be SÚRAO, and the safety of the DGR is verified by SÚJB in the framework of the administrative procedures for issuing the various licenses.

Verification and validation of computational programs is one of the basic requirements of the SÚJB for safety assessment. Researchers from many research organisations and universities are involved in the development, verification and validation of computing programmes. In addition, the issue of long-term safety prediction is addressed in practically all countries operating nuclear power. Mutual exchange of knowledge also contributes to the identification of possible errors and more accurate predictions.

A natural analogue is the occurrence of a type of process (material) in the natural environment that is similar to, or has some relationship to, processes (materials) that may occur in the actual deposit or its near or distant surroundings. Natural analogs are programs that evaluate natural interactive processes over long "geologic" time periods. The natural process interactivity extends from the time of origin of the environment under study to the present. The advantage of natural analogs is the long-term nature of the interactive process, which cannot be achieved in an underground laboratory. The results from natural analogue studies are mainly applied in the field of disposal facility safety verification through validation of interactive process models and in the public engagement programme.

One of the basic principles of safety assessment is to use a conservative approach, which means that the outcome of the assessment includes its least favourable, plausible variants. The prediction of long-term safety should therefore always be less favourable than the reality may be (answer in cooperation with SÚRAO).

3. I don't believe that you can prove the safety of the DGR for thousands of years in advance. How can government officials (from Prague...) be sure?

The post-closure safety assessment of the DGR is burdened with many uncertainties. Therefore, a conservative approach is used in the safety assessment, i.e. the least safety-appropriate initial assumptions and input data are used. The safety analyses of the HLS, which are part of the documentation submitted by the prospective licensee to the SÚJB, include sensitivity analyses and probabilistic assessments that quantify the effect of uncertainties on the overall safety of the HLS. Using these methods, the safety of the DGR can be demonstrated with reasonable confidence over time horizons on the order of tens to hundreds of thousands of years after closure of the DGR.

In the Czech Republic, experts from many research organisations and universities are involved in the evaluation of disposal facility safety.  This issue is being addressed with the participation of Czech researchers in a number of international projects. The results of the evaluation are discussed at international seminars and are checked by experts from the SÚJB and subsequently by experts from the International Atomic Energy Agency (IAEA).  The European Union supports this issue in the framework of the EURATOM programme, where sub-issues are addressed jointly by experts from different countries. The results achieved are also discussed in the framework of bilateral cooperation with Finland and Sweden, which are much further along in the preparation of the DGR than the Czech Republic. Close cooperation with Germany is also beginning to develop. SÚRAO is a management organisation that coordinates research activities in the Czech Republic to prioritise those areas that correspond to the needs of DGR preparation and summarises the partial results of experts from different disciplines in the form of safety documentation, which is submitted to SÚJB for a decision on whether all safety requirements are met (response from SÚRAO).

4. Where are the results of conservative estimates of exposure to a representative person available? How was the representative person identified?

The results of the safety analyses are presented both in SÚRAO documents and in the outputs of SÚJB-funded research projects. The general procedure for the identification of a representative person, i.e. an individual from the population representing a model group of individuals who are, for example, most exposed to radiation from DGR, is presented in Annex 5 of Decree No. 422/2016 Coll.

For all of the original nine potential DGR sites, SÚRAO has prepared safety case studies in 2018, which include estimates of exposure to a representative person based on preliminary information about the sites and the DGR project. For all sites, this exposure, expressed as the so-called effective dose rate, is well below the optimization limit of 0.25 mSv/year.

Similarly, in 2015-2018, the SÚJB awarded the project Development of a Deep Geological Repository by public procurement. The objective of the project was to independently assess the safety of sites proposed for future DGR siting based on input provided by the future disposal facility operator in the scope of the tender safety report for the DGR siting permit. The results for the hypothetical DGR site confirmed the conclusions of the safety report studies.

5. What will be the impact of deep disposal on the radiation situation in the environment? Will dose rates ("radiation") increase in the surrounding area? By how much?

The impact of the operation and the period after the shutdown of the DGR on the radiation situation near the DGR will be practically unmeasurable. Only in the event of a radiological emergency during operation of the DGR could there be a limited release of radioactive substances into the environment.

Assuming a breach in the leak-tightness of the fuel assemblies, the releasable fractions of radionuclides contained in the fuel, i.e. a fraction of the total radionuclide activity, may be released. The HAzardous Radioactivity Propagation (HARP) software, which is also used at Temelín Nuclear Power Plant, is used to simulate the transport of radionuclides through the air, taking into account the dose from external exposure as well as the effective dose rates from ingestion (food chains) and inhalation. Preliminary calculations show that for a worst-case event, the effective dose for an adult individual will not exceed the unit microSv/year and is thus well below the optimisation limit of 0.25 mSv/year, while at the same time no radiological emergency will occur (answer in cooperation with SÚRAO).

6. What will be the radiation situation (dose rates) in the deep repository? What will be the doses to the workers/services of the DGR (e.g. compared to a nuclear power plant, health care...)?

The radiation situation in the DGR will be comparable to the radiation situation in operating nuclear power plants and SNF storage facilities and will fully comply with the legislative requirements specified in the Atomic Act and Decree No. 422/2016 Coll.

During the years 2017 - 2019, the maximum annual effective dose rate at the EDU was 2.4 mSv/year and during the years 2010 - 2019 at the SF storage facilities on the EDU site was 0.5 mSv/year. Although it is not possible to determine the exact effective dose rates of the DGR workers at present, it can be assumed that the annual effective dose rate of the DGR workers, when handling SF in the DGR (receiving of transport and storage packages, loading of SF to disposal packagings, transport and settling of disposal packages in the underground areas of the DGR), will not exceed the values indicated for NPP Dukovany (or NPP Temelín) workers.

7. What hydraulic model was used to assess the propagation of radionuclides released from the disposed RAWs to the environment where they will be disposed?

Within the framework of research works dealing with the study of radionuclide propagation between individual barriers of the DGR, more precisely the propagation from the disposal packaging to the damping material (bentonite) and subsequently from the bentonite to the rock environment, the flux from the so-called "contaminant transport" was modelled in the GoldSim - Contaminant Transport Module. The "near-field interaction" to the rock environment was simulated using discrete fractures, where one fracture represents the fracture intersecting the containment borehole and the other fracture intersects the access corridor.

The flow rate from the bentonite into the fracture that intersects the disposal well is characterized in the model by diffusive flow with an initial zero concentration of radionuclide in the water flowing through the fracture.

For a fissure crossing a corridor, the volumetric velocity is equal to the flow velocity in the faulted rock around the corridor. This modelling policy has been tested in a safety analysis study to assess the long-term safety of a deep geological repository.

Since disposal package and engineering barriers are considered the same across all sites, only the hydraulic and transport model in the remote interaction region (i.e., in the rock environment itself) was used to evaluate the sites, which accounts for differences in hydrogeologic conditions between sites. This model describes the flow and movement of radionuclides from the engineered barrier/rock environment interface to the contact between the rock environment and the biosphere (response from SÚRAO).

8. What kind of DGR safety research is performed.

The research for the purpose of the DGR safety proof is provided by SÚRAO, which is responsible for ensuring the safety of the DGR.  Substantive support for the safety assessment, such as the development, verification and validation of computer programmes, determination of input data and obtaining other information for the safety assessment, is provided by research organisations and universities on the basis of public tenders. Between 2014 and 2020, SÚRAO managed a research project aimed at supporting the safety assessment of DGR, which involved more than 200 researchers from more than 10 research organisations and universities. The project produced more than 200 interim and final research reports in both Czech and English. A follow-up project is currently under preparation to further deepen our knowledge, develop, verify and validate mathematical models, and obtain input data and other information needed to reduce uncertainties in disposal facility safety assessment (response from SÚRAO).

9. Why are underground labs being built?

The underground laboratories are built to verify the behaviour of the rock environment at the depth of the disposal facility and to develop the disposal policy. They will allow to specify how waste can be disposed of in a future HLF, to test all engineering components of the HLF and to predict the long-term behaviour of the HLF, in particular of the host rock environment. It is also used to validate new optimal mining technologies and to verify facts predicted by surface exploration methods and drilling. This knowledge cannot be obtained in any other way, without carrying out the relevant work at the depth of the planned disposal facility (response from SÚRAO).

10. Will there be an "emergency" zone in the vicinity of the DGR? What will be its size? Will there be warning devices (sirens, etc.) near the DGR?

The establishment of an emergency planning zone is based on its definition in Section 4(1)(k) of Act No. 263/2016 Coll. and the requirements of Annex 1, Part 1 of Act No. 263/2016 Coll. According to Decree No. 359/2016 Coll., Annex 2(a), an emergency planning zone is not established if the frequency of occurrence of a radiation accident is less than 10/year^-7. Therefore, it can be considered highly probable that an emergency planning zone will not be established for the HP in the future.

However, if an emergency planning zone is established, then it is necessary to place a set of representative radiation monitors in it. The legislation explicitly requires the measurement of H*(10) and the air's volumetric activity (selected radionuclides, initially ⁸⁵Kr, later deposited radionuclides).

According to "Risk Informed Support of Decision Making," (SÚJB, 2008), p. 26, it is impractical to operate monitoring at distances greater than 25 km. This material primarily addresses the issue of emergency planning zones for nuclear power plants - in the case of a deep geological repository, the maximum distance of monitoring points will be less.

A common approach is one metering point per municipality/residence. For example, the NPP Dukovany telemetry system consists of two circuits. The monitors of the first circuit are located around the perimeter of the NPP Dukovany site. The monitors of the second system are in the surrounding municipalities. With the exception of one monitor in Moravský Krumlov (due to the higher population), all of them are located within 5 km of the EDU, so that all main directions of possible radioactive material spread are covered.

The monitoring stations must be equipped with very sensitive detectors and transmit data immediately to the radiation control room or equivalent, otherwise there is no hope of detecting a leak in time. The operation of stationary monitors must be ensured under all foreseeable weather conditions (detection, data processing, communication with the monitoring station, power supply) (response from SÚRAO).

11. Where are the results of analyses of the impacts of hypothetical radiation incidents during transport, handling and storage of spent fuel available? What are the conclusions of these analyses? Who carried out these analyses?

The results of the radiation extraordinary event (RMU) analyses for transport, handling and storage of SNF are presented in the documentation for permitted activities submitted to the SÚJB as part of the administrative procedures for the issuance of the relevant permits and submitted by the applicant for the permitted activity or for the approval of the packaging type. The conclusions of these analyses provide evidence of safety in the management of SF in accordance with Act No. 263/2016 Coll. and its implementing regulations.

The documentation for the permitted activities includes safety reports that analyse all considered operational conditions and design events during transport and storage of SF (see Annex 1, section 1, point 4, letter f) and section 4, letter d) of Act No. 263/2016 Coll.). In the case of the use of transport and storage packages, these analyses are based on the package type approval documentation, in which a demonstration of the fulfilment of all safety functions of the package is provided (see Annex 2 of Act No. 263/2016 Coll.). Package type approval may be applied for by its manufacturer, importer, distributor or any other person who demonstrates a legal interest in the type approval of the product (e.g. its user, see Article 138(1) of Act No. 263/2016 Coll.) and who is also responsible for the content of the submitted documentation.

The results of the analyses prove the safety of SF management in accordance with Act No. 263/2016 Coll. and its implementing regulations, in particular Decree No. 379/2016 Coll. In the event of a large transport aircraft crash on the Temelín SNF building, for example, there is no mechanism that could achieve the containment of the stored CASTOR 1000/19 type cask. A similar conclusion applies to the SF storage facilities at the NPP Dukovany site.

Analyses and evaluations of RMUs during transport of SNF at nuclear power plant sites have shown that scenarios leading to radiation accidents and thus the need to introduce urgent measures for the protection of the population can be excluded. None of the considered accident scenarios leads to such destructive effects on the RPU that could break its leak tightness and lead to the release of the inventory of releasable radionuclides from the SNF into the environment.