In-situ monitoring of micro-tunnel geometry change based on distributed optical fiber strain sensing

Underground radioactive waste disposal facilities must be monitored to guarantee their correct and safe exploitation during the early stage of their service life. This will be the case in Cigéo, the french industrial geological disposal facility for high-level and intermediate-level long-lived waste, which must be monitored for the first a hundred years at least. For this purpose optical fiber strain sensing cables can be chosen for their low intrusivity and resistance to harsh radiation environment. In this paper, a monitoring method able to follow shape evolution of the structure's section, based on distributed strain sensing is tested on a high-level waste disposal cell mock-up, a steel lined micro-tunnel of 762~mm of outer diameter. The cell demonstrator is built in Meuse/Haute-Marne Underground Research Laboratory, in the same geological environment as the one envisaged for Cigéo but it is empty of waste. After 7 months of loading of the rock around the mock-up, a maximum of 2 mm of diameter reduction is observed with Rayleigh-scattering based strain sensing technique and 10 mm of spatial resolution around the circumference. This validation under realistic test conditions opens up to the use of the method in-situ for tunnel monitoring in harsh environment.

Modeling and simulation of ductile damage in forged steel used to encapsulate high-level radioactive waste

Andra, the French national radioactive waste management agency, is in charge of studying the disposal of high-level and long-lived intermediate-level waste (HLW and ILW-LL) in a deep geological repository. According to the reference concept, it is planned to encapsulate high-level waste in non-alloy P285NH steel overpacks before inserting them into horizontal steel cased micro-tunnels. This work is a part of the study about the long-term behavior of a welded steel overpack subjected to external hydrostatic pressure and several localized loading paths. Indeed, the main objective of this work is to develop the most suitable model for non-alloy steel P285NH to be used in the prediction of the long-term overpack behavior. Dealing with a ductile steel, elastoplastic constitutive equations accounting for mixed nonlinear isotropic and kinematic hardening strongly coupled with ductile isotropic damage are adopted. They are formulated based on the classical thermodynamics of irreversible processes framework with state variables at the macroscopic scale, (Germain, 1973) (Lemaitre 1985, Saanouni 2012). In this paper, a new coupling formulation between the scalar isotropic ductile damage and the deviatoric and spherical part of the Cauchy stress and elastic strain tensors is proposed. In order to calibrate the developed model on P285NH steel, multiple tensile tests are performed using classical cylindrical specimens, notched specimens and double notched specimens. In the last part, some experimental fields are measured using digital image correlation. Application is made to a simplified overpack represented by thick walled cylinder subject to compressive loading path. A FEM (Finite Element method) crushing operation of an overpack’s cylindrical part has simulated and analysed.

Radioactive Fission Waste from Molybdenum-99 Production and Proliferation Risks

Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.

Site selection for the best clay-hosted repository in Switzerland

The site selection process for the deep geological repositories in Switzerland is in its final phase. All three remaining sites fulfil the requirements of the repositories for low level waste (LLW) as well as for high level waste (HLW) and spent fuel. Using surface-based exploration methods, including 3D seismic studies and deep bore holes, Nagra has recently collected the necessary data to select the most suitable site. The sites will be compared based on 13 technical criteria defined in 2008 and already used in the 2 previous phases of the selection process. The criteria comprise safety-related aspects including the barrier properties and their long-term stability as well as the construction suitability of the repository and its access facilities. If one site offers optimal conditions for both HLW and LLW and the space requirements can be met at that site, a co-disposal facility will be proposed. This facility will then comprise separate emplacement areas with specific safety concepts for the different waste categories. In our contribution we present the overall approach for the surface-based exploration in the different stages of the site selection process. We show how we integrated third party data, seismic surveys, surface mapping as well as deep and shallow bore holes to drive the site selection process. Example data sets from the ongoing deep drilling campaign (clay content, hydrotest data, porewater isotope profiles etc.) and recent 3D seismic surveys are used to illustrate our approach to distinguish the remaining sites according to long-term safety and to underline the large safety margin of the selected clay host rock in long-term evolution scenarios.

Cement as a component of the multi-barrier system: radionuclide retention in cementitious environments

Safety concepts regarding nuclear waste disposal in underground repositories generally rely on a combination of engineered and geological barriers, which minimize the potential release of radionuclides from the containment-providing rock zone or even their transport into the biosphere. Cementitious materials are used for conditioning of certain nuclear waste types, as components of waste containers and overpacks, as well as being constituents of structural materials at the interface between backfilling and host rock in some repository concepts. For instance, the preferred option for the disposal of high-level waste (HLW) in Belgium is based on the supercontainer design, which consists of a carbon steel overpack surrounded by a thick concrete buffer (Bel et al., 2006). In the event of formation water interacting with cementitious materials, pore water solutions characterized by (highly) alkaline pH conditions will form. These boundary conditions define the chemical response of the radionuclides, but also influence the behaviour of neighbouring components of the multi-barrier system, e.g. bentonitic or argillaceous backfilling and host rock. Hardened cement paste or Sorel cement are considered to be main sorbing materials present in the near field of repositories for low- and intermediate-level waste (L/ILW). Hence, interactions of radionuclides with cementitious materials represent a very important mechanism retarding their mobility and potential migration from the near field (Wieland, 2014; Ochs et al., 2016). While the quantitative description of the sorption processes (usually in terms of sorption coefficients, i.e. Kd values) is a key input in the safety analysis of nuclear waste repositories, detailed mechanistic analysis and understanding of sorption phenomena provide additional scientific arguments and important process understanding, and thus enhance both the quality of safety arguments and the overall confidence in the safety assessment process. Research at KIT-INE dedicated to the interaction of cementitious materials with radionuclides is conducted in the context of different repository concepts, including clay (low- and high-ionic strength conditions), crystalline rock or rock salt. Experimental and theoretical studies are performed within the framework of national (GRAZ, BMWi) and international (CEBAMA and EURAD-CORI, EU Horizon 2020 Programme) projects, extending to third-party projects with several waste management organizations in Europe, e.g. SKB (Sweden), ONDRAF-NIRAS (Belgium) or BGE (Germany). The combination of classical experimental (wet chemistry) methods, advanced spectroscopic techniques and theoretical calculations provides both an accurate quantitative evaluation and a fundamental understanding of the sorption processes. Examples of recent studies at KIT-INE on radionuclide behaviour in cementitious systems in the context of both L/ILW and HLW will be presented in this contribution to explain methodologies, scientific approaches and results. The present state of knowledge as well as main remaining uncertainties affecting the retention processes of radionuclides in cementitious environments under different conditions will be critically discussed, also in view of current international research activities and repository projects.

A citizens workgroup helps researchers reflect on their work

In our transdisciplinary project (http://www.transens.de) several academic disciplines work on questions and solutions for high-level nuclear waste disposal in Germany. Adding to this interdisciplinary setting, in our sub-project on trust, we have enlisted a group of 16 citizens (citizens workgroup, CWG), reflecting with us on our research with regard to research questions and approaches. In this talk, we present results from a joint workshop of researchers and the CWG on the role of trust in scientists. We want to understand how participants perceive and relate to science and how this may affect trust in scientists and experts. During a first workshop (conducted online in March 2021), the CWG members were allocated to three breakout groups (1)–(3) to discuss three guiding questions: a. What expectations do you have of a scientist from Social Sciences and Humanities (SSH)/Natural Sciences and Engineering (NSE) in the TRANSENS project? b. What characterizes a scientist from SSH/NSE that you trust? c. What would have to happen for you to no longer trust him/her? The group in breakout room (1) was asked to answer questions (a)–(c) for a fictitious SSH researcher. The group in breakout room (2) was asked to do the same for a fictitious NSE researcher. The group in breakout room (3) was asked whether they perceived differences in the trust of SSH and NSE researchers. For SSH we found that a scientist should be sympathetic and non-condescending, represent a neutral point of view, and consider all opinions. Remarkably, discussants in this group struggled to define a role for SSH scientists in high-level waste disposal research. Some participants ascribed SSH a moderating or mediating role. If necessary (e.g. communication of results), mediation between the NSE and the public can be added. SSH scientists may train other scientists with regard to their performance. Participants stated that scientists from NSE should present information in full and clarify the current state of research (provisional nature of knowledge acknowledged). A scientist should not conceal any information, uncertainties, or risks, be neutral and objective, have experience in the field (professionalism, experience, reputation), and not exclusively reproduce one's own opinion or spread untruths. In all groups, participants judged the NSE as “harder”, more serious and more tangible. It was also discussed that the scientist's appearance is of great importance and that a scientist can “pick up” the audience in an exciting way when communicating. Personal experience was mentioned several times in all groups as a basis for trust. These results match findings on the effect of value similarity (Siegrist et al., 2000, and own survey in 2020 in Germany): If a person is perceived as advocating similar values, it is more likely that I trust that person. Personal experience of – among other issues – similar values may increase trust. Moreover, and importantly, trust also emerges when participants know exactly how issues will be considered, e.g. if input from the CWG is considered or not – and if not: why? In general, participants want to be taken seriously. Thus, transparent and binding rules for all participants may be key for a relationship of trust.

Spectroscopic and chemical investigations on volatile fission and activation products within the fuel-cladding interface of irradiated pressurised water reactor fuel rod segments

In Germany, the present waste management concept foresees the direct disposal of spent nuclear fuel (SNF) in deep geological repositories for high-level waste available by 2050, at best. Until then, SNF is encapsulated in dual-purpose casks and stored in dry interim storage facilities. Licenses for both casks and facilities will expire after 40 years following loading of the cask and emplacement of the first cask in the storage location. Yet, due to considerable delays in the site selection process and the estimated duration for construction and commissioning of a final repository of at least 2 decades, a prolonged dry interim storage of SNF is inevitable (ESK, 2015). Concerning these considerable timespans, integrity of the cladding is of utmost importance regarding the ultimately conditioning of the fuel assemblies for final disposal. Various processes strain the structural integrity of Zircaloy cladding during reactor operation and beyond such as delayed hydride cracking, fuel-cladding chemical interactions or irradiation damage induced by α-emitters present in the fuel pellet's rim zone (Ewing, 2015). Especially with higher burn-up, the gap between fuel and cladding closes and results in the formation of an interaction layer, in which precipitates of fission and activation products are present, displaying an interface for degradation processes. For chemical analysis and speciation of these agglomerates, Zircaloy-4 and SNF specimens were sampled from fuel rod segments irradiated in commercial pressurised water reactors during the 1980s. Zircaloy-4 specimens were taken from an UOX (50.4 GWdtHM-1) and mixed oxide fuel (MOX) (38.0 GWdtHM-1). In addition, SNF fragments were sampled from the closed gap of both fuel types to examine volatile activation and fission products, which had been segregated from the centre to the pellet periphery during irradiation and thus contribute to the possible chemically assisted cladding degradation effect of the precipitates within the fuel-cladding interface. Spectroscopic analysis of precipitates within the interface layer between fuel and cladding were performed by optical microscopy, X-ray absorption and X-ray photoelectron spectroscopy, as well as by energy-dispersive scanning electron microscopy. Moreover, the radionuclide inventory of the respective Zircaloy-4, fuel and interaction layers was determined using liquid scintillation counting, γ-spectroscopy, gas mass spectrometry, ion chromatography and inductive-coupled plasma mass spectrometry and compared to results received by MCNP/CINDER and webKORIGEN calculations. In this study, we provide results regarding the speciation and chemical composition of previously identified Cs-U-O-Zr-Cl-I bearing compounds found in the interaction layer of irradiated nuclear fuel and inventory analyses of radionuclides present therein, with particular emphasis on Cl-36 and I-129. Furthermore, the agglomerates within the fuel-cladding interface were characterised for the first time utilising synchrotron radiation-based Cl K-edge and I K-edge measurements, resulting in compounds with structural similarities to CsCl and CsI. The outcomes obtained from this study provide further insights into the complex chemistry within the fuel-cladding interface with respect to the aging management and integrity of SNF under the conditions of interim storage. In future studies we will examine whether the different compounds at the fuel-cladding interface have the potential to affect the mechanical properties of Zircaloy cladding.

Deep borehole disposal of intermediate-level waste

Around the world, deep borehole disposal is being evaluated for intermediate-level waste (ILW), high-level waste (HLW), spent nuclear fuel (SNF), separated plutonium waste and some very high specific activity fission product waste. In Australia, long-lived ILW from research reactors and radiopharmaceutical production represents the principal waste stream that requires deep geologic disposal. Whilst the Australian government has not yet made a decision on its preferred strategy for ILW disposal, deep borehole disposal of small volumes of ILW would be a more cost-effective and modular solution compared to a conventional geologic disposal facility (GDF). CSIRO, ANSTO and SANDIA have created an international partnership to execute a full-scale borehole research, development and demonstration (RD&D) project in Australia. The project will demonstrate the technical feasibility of the long-term safety of borehole disposal in deep geological formations. The execution of this project could also demonstrate options for nuclear waste disposal that would reduce proliferation risks, potentially up to the termination of compliance with international safeguards requirements. The RD&D includes demonstration of surface handling and waste/seal emplacement capabilities, basic research on foundational science areas, and full-scale field testing in both a deep characterization borehole and a larger-diameter (0.7 m or 27.5 inch) 2000 m deep demonstration borehole. The multi-barrier system designed for such a deep disposal borehole concept places much less reliance on engineered barriers at the disposal zone to achieve safety as compared to a conventional GDF. It rather relies on geological features for waste containment. The concept being explored uses disposal containers with primary waste packages, such as vitrified waste canisters, inside; to be both cost effective and fit for purpose, such a container could have a mild steel-based structural component with copper coating. A critical review of six coating technologies showed that cold spray has the greatest advantages, such as minimal porosity and compressive residual stress. The RD&D has delivered novel enabling tools that assist with site screening, borehole design and post-closure safety assessments. For instance, an automated geological fault mapping and meshing tool was developed that assists with ranking the suitability of potential disposal sites based on proximity to faults. New codes were developed for better representation of fault zones in 2D/3D numerical flow and transport models, while also being more efficient to execute. Post-closure safety assessments tested the sensitivity of long-term safety with respect to disposal depth, rock permeability and sorption. Heat transport calculations explored the sensitivity of temperature evolution within the borehole to parameters such as heat load, borehole depth, geothermal gradients and rock thermal conductivity. For verification of host rock tightness while also demonstrating the absence of recent groundwater, a new noble gas analytical facility has been established for measuring rare noble gases in mineral fluid inclusions as indicators of very old pore fluids.