Two-phase transport in a cemented waste package considering spatio-temporal evolution of chemical conditions

The long-term evolution of cemented waste packages is governed by (bio)chemical reactions between waste, cement, and barrier materials, and by transport processes inside the waste package and through openings in the waste package hull. Inside a waste package, gases can be generated by pH-dependent anoxic corrosion of metals and the degradation of organic matter. The (bio)chemical reactions consume water and will not proceed under dry conditions. The degradation of cementitious materials lowers the alkaline pH of the pore water. The modeling of such a complex feedback system shows that the internal structure of a waste package and the exchange of mass with the environment (boundary conditions) are major factors that determine the importance of process couplings. The (bio)chemical reactions are controlled by internal re-distribution of water predominantly via vapor transport. Calculated gas generation rates were found to be affected by dry-out processes inside the drum.

Post-Closure Performance Assessment for Deep Borehole Disposal of Cs/Sr Capsules

Post-closure performance assessment (PA) calculations suggest that deep borehole disposal of cesium (Cs)/strontium (Sr) capsules, a U.S. Department of Energy (DOE) waste form (WF), is safe, resulting in no releases to the biosphere over 10,000,000 years when the waste is placed in a 3–5 km deep waste disposal zone. The same is true when a hypothetical breach of a stuck waste package (WP) is assumed to occur at much shallower depths penetrated by through-going fractures. Cs and Sr retardation in the host rock is a key control over movement. Calculated borehole performance would be even stronger if credit was taken for the presence of the WP.

Radiation Protection Calculations for the New Radioactive Waste Interim Storage Facility of NCSR “Demokritos"

The present study concerns the determination of the maximum acceptable contact dose rate per radioac- tive waste package for safekeeping at the New Radioactive Waste Interim Storage (NRWIS) of the National Centre for Scientific Research “Demokritos” (NCSR “D”). The NRWIS facility is used for temporary storage of spent/ orphan sealed sources, devices like lightning rods and primary radioactive waste. The contact dose rate per package is determined in a level that even in case the highest radiation background is built up in- side the storage facility, the doses to the workers will not exceed the maximum permissible doses. The total dose that a worker receives inside the facility should not exceed one half of the annual occupational dose constraint of 6 mSv. Furthermore in cases of the highest radiation background inside the facility, shielding calculations are performed.

Consequence modelling of hypothetical post-closure criticality events for spent fuel disposal

In support of the Radioactive Waste Management (RWM) safety case for a geological disposal facility (GDF) in the UK, there is a regulatory requirement to consider the likelihood and consequences of nuclear criticality. Waste packages are designed to ensure that criticality is not possible during the transport and operational phases of a GDF and for a significant period post-closure. However, over longer post-closure timescales, conditions in the GDF will evolve.For waste packages containing spent fuel, it can be shown that, under certain conditions, package flooding could result in a type of criticality event referred to as 'quasi-steady-state' (QSS). Although unlikely, this defines a 'what-if' scenario for understanding the potential consequences of post-closure criticality. This paper provides an overview of a methodology to understand QSS criticality and its application to a spent fuel waste package.The power of such a hypothetical criticality event is typically estimated to be a few kilowatts: comparable with international studies of similar systems and the decay heat for which waste packages are designed. This work has built confidence in the methodology and supports RWM's demonstration that post-closure criticality is not a significant concern.

Understanding the likelihood and consequences of post-closure criticality in a geological disposal facility

A geological disposal facility (GDF) will include fissile materials that could, under certain conditions, lead to criticality. Demonstration of criticality safety therefore forms an important part of a GDF's safety case.Containment provided by the waste package will contribute to criticality safety during package transport and the GDF operational phase. The GDF multiple-barrier system will ensure that criticality is prevented for some time after facility closure. However, on longer post-closure timescales, conditions in the GDF will evolve and it is necessary to demonstrate: an understanding of the conditions under which criticality could occur; the likelihood of such conditions occurring; and the consequences of criticality should it occur.Work has addressed disposal of all of the UK's higher-activity wastes in three illustrative geologies. This paper, however, focuses on presenting results to support safe disposal of spent fuel, plutonium and highlyenriched uranium in higher-strength rock.The results support a safety case assertion that post-closure criticality is of low likelihood and, if it was to occur, the consequences would be tolerable.

Thermal dimensioning to determine acceptable waste package loading and spatial configurations of heat-generating waste packages

Heat-generating waste provides a number of additional technical challenges over and above those associated with the disposal of ILW. A priority area of work for Radioactive Waste Management (RWM) concerns the effect of heat on the engineered barrier system, and how this may be mitigated through the management of heat (thermal dimensioning) in a UK Geological Disposal Facility (GDF). The objective of thermal dimensioning is to provide a strategy to enable acceptable waste package loading and spatial configurations of the packages to be determined in order to enable high-heat generating waste to be successfully disposed in a GDF. An early focus of the work has been to develop a thermal modelling tool to support analyses of different combinations of package assumptions and other GDF factors, such as spacing of those packages, to assess the compliance with thermal limits. The approach has a capability to investigate quickly and efficiently the implications of a wide range of disposal concepts for the storage of spent fuel/HLW and the dimensions of a GDF. This study describes the approach taken to undertaking this work, which has included a robust appraisal of the key data (and the associated uncertainty); recent thermal dimensioning analysis has been performed to identify constraints on those disposal concepts.