Many properties need to be considered in waste form development. Mechanical properties are important from the point of view of material integrity in a storage or disposal environment. Of particular importance is the long-term chemical durability of the waste form as this will influence the release of toxic elements due to leaching under disposal conditions. Chemical durability can also change over very long timescales depending on the radiation stability of the waste form, particularly as this may adversely affect the dissolution rates through, for example, the generation of corrosive radiolysis products in the leaching solution.

Durability studies performed using a modified MCC-1 procedure on AWE monolithic material which had undergone accelerated ageing trials through the substitution of 238Pu for 239Pu demonstrate this adverse effect. 238Pu-doped material which had been aged for 1820 days (total a radiation fluence of 4 x 1018g-1) was leached in water at 40°C for 28 days and com­pared with 239Pu-doped material (unaged) and the results are summarized in Table 25.8. It can be seen that the release of elements from the aged 238Pu

samples is considerably greater than that from the unaged 238 or 239 samples.

Internationally, several different systems have been developed for the clas­sification of radioactive wastes. These are generally based on half-life, activ­ity levels, origin or source, or the degree of isolation required. In general, low level waste (LLW) contains radionuclides with low activities and short half-lives and generates no heat; some systems differentiate a subcategory for very low level waste (VLLW). Intermediate level waste (ILW) may contain radionuclides with low to intermediate activities and short to long half-lives, generating no to negligible heat. High-level waste (HLW) con­tains radionuclides with high activities, long or short half-lives or both, and generates heat (Rempe, 2007).

In Germany, as mentioned earlier, the BMU defines nuclear waste for disposal purposes based on its heat generating capacity, as either waste with negligible heat generation or heat-generating waste. In the German system waste with negligible heat generation consists of VLLW, LLW and ILW, while waste classified as heat generating consists of both spent nuclear fuel (SNF) and HLW. In accordance with federal policy as promulgated by the BfS, both waste types are to be disposed of in waste-specific deep geological repositories (BfS, 2011c). Construction, operation and closure of a reposi­tory must be approved according to the Atomic Energy Act (AtG §9b) as part of a planning approval process.

The major sources of radioactive wastes in Germany are associated with nuclear fuel cycle activities, power generation, research facilities, the re­importation of HLW associated with the reprocessing of SNF in the United Kingdom and France, decommissioning of the various nuclear facilities, and the use of radioisotopes in medical, research and industrial applications. Other materials, primarily associated with the decommissioning of nuclear facilities, which are either not radioactive or only weakly radioactive, can be released from nuclear regulatory control by permit providing applicable regulatory conditions are met (Chapter 2, Section 9, §29 of StrlSchV). The BfS estimates that a total of approximately 290,000 m3 of waste with negli­gible heat-generating capacity will require disposal (BfS, 2011d). Of these, approximately 161,000 m3 of the waste is expected from decommissioned NPPs by 2080 (BfS, 2011c). The current inventory of heat-generating nuclear waste requiring geological disposal in Germany as of 31 December 2010 is given in Table 14.1.

Absence of those responsible

Polluted sites are predominantly from the radium mining industry. The industry died out in the 1920s after a boom period in the wake of the work of Marie Curie and medical applications implemented during the First World War. Thus the last radium extraction site stopped in 1928 (in the town of Nogent-sur-Marne). It is thus unrealistic to look for any responsible body still in place (there is now a 30-year proscription applicable to industrial activities following the decision of the State Council on the ‘Allusuissse’ issue). The situations ANDRA inherits are therefore situations from the national industrial heritage, to be managed as best as possible.

17.1.1 Devolution of UK governmental powers

The Scotland Act received Royal Assent on 19 November 1998 and the Scottish Parliament sat for the first time on 6 May 1999 (White and Yonwin,

2004). This act transferred powers for specific issues to the Scottish govern­ment and reserved powers with the UK government for the remainder. The devolution of powers to the Scottish government for specific issues and not others has an influence on the areas of nuclear power, RAW management and the environment.

Among the reserved powers, the UK government controls energy, which includes electricity and nuclear energy, and also defence and national secu­rity. Among the devolved powers, the Scottish Parliament has health, plan­ning and the environment. This means that nuclear installations in Scotland are subject to legislation in specific areas from the UK government and in other areas from the Scottish government. They are regulated by some agencies which report to the UK government and by others that report to the Scottish government.

Scotland is similar to the rest of the UK in its hospitals and industries using radioactive sources for medical and industrial purposes. The use of these sources is controlled by the suppliers who in most cases are also responsible for their storage or disposal after use. There are many movements of these radioactive sources daily under controlled conditions and in authorised containers.

17.2 Conclusion

Although the volumes of radioactive waste and decommissioning activities in Scotland are small compared with the total UK liabilities, they are nev­ertheless diverse and challenging. Dounreay is the second most challenging site in the UK after Sellafield. Scottish radioactive waste managers and nuclear site operators manage their responsibilities both within UK require­ments and legislation and the Scottish government’s specific policies on RAW management. Although Scotland has significantly different approaches to some aspects of HAW management, these are not creating operating problems at present. During the next few decades of develop­ment of disposal technologies in the UK as a whole, and the Scottish gov­ernment’ s commitment to review its HAW policy every ten years, closer alignment and coordination are possible.

Waste disposition for commercial medical isotope production

The DOE/National Nuclear Security Administration (NNSA) is working to accelerate commercial production of the medical isotope molybdenum-99 (Mo-99) in the United States without the use of highly enriched uranium (HEU). Mo-99 ’s primary uses include the detection of disease, including heart disease and cancer, and the study of organ structure and function. The isotope’s short half-life and excellent binding properties make it uniquely suited for medical procedures. However, its 66-hour half-life prevents it from being stockpiled during periods of shortage. Mo-99 is a crucial radio­isotope used in approximately 80% of all nuclear medicine diagnostic pro­cedures and in roughly 50,000 diagnostic and therapeutic nuclear medicine procedures performed every day in the United States.

In cooperation with commercial partners and the US national laborato­ries, DOE/NNSA is supporting the US private sector in developing inde­pendent, non-HEU-based technical pathways to produce Mo-99 in the United States by 2014. The NRC or Agreement State would have to license any new commercial production facility. The expected waste streams from the production of Mo-99 are likely to include radioactive waste for which there is currently no commercial disposal path. The projects are under development, and production has not yet commenced at the time this book was written. However, disposition of specific waste and spent nuclear fuels and targets resulting from Mo-99 production could impact the technical and economic viability of each of the projects. Until a disposal path is identified, producers of this medical isotope would need to provide onsite storage.

Currently, a key focus area for the LLRWMO is putting in place a strategy to address sites in Canada’s north that were contaminated, long ago, by the
spillage of radioactive ores in transport. The contamination is located along what is known as the northern transportation route as shown in Fig. 19.4 (LLRWMO, www. llrwmo. org), a 2200 km route beginning at the former Port Radium site in the Northwest Territories and extending to northern Alberta. The LLRWMO is adapting methods that it has successfully used in Canada ’s southern regions. These methods of community engagement and technical approaches take into account the geography and the environ­ment, while respecting the ways of inhabitants in the north. The success of the LLRWMO in southern communities has been based on building confi­dence with the communities involved through a carefully designed process, including cultivating stakeholder involvement early on in the process. Building and maintaining a community’s confidence require constant com­mitment, significant resources and mutual effort. In northern communities, these engagement processes include taking into account traditional and local knowledge and providing training and

19.4

policy decision so that the community can participate in the stewardship of their natural environment. One of the challenges of such projects built on a participatory approach is the need to find and achieve a balance between stakeholder representation, stakeholder participation, and project progress and implementation. That balance can vary with the individual project and its stakeholders.

J.-I. Y U N, Y. H. J E O N G and J. H. KIM, KAIST, Korea DOI: 10.1533/9780857097446.2.673

Abstract: Republic of Korea currently operates 21 nuclear units providing one-third of the nation’s electricity. Low and intermediate level radioactive materials emanating from these plants, medical facilities, research reactors, and industry need to be safely stored and managed. Disposal of spent nuclear fuel is also an important national issue. This chapter reviews the current state of affairs in Korea and examines the national policy, strategy, and direction for managing spent fuel and radioactive waste (RAW) materials. Decontamination of waste materials is also discussed.

Key words: Republic of Korea, radioactive waste (RAW), spent nuclear fuel (SNF) storage, disposal, decommissioning, decontamination.

21.1 Introduction

The twenty-first century’s grand challenges are aptly characterized by energy, environment, and economy — the so-called tri-lemma of sustainabil­ity. These three Es are intricately interconnected, and balancing them is necessary for a healthy society. Many of this century’s issues are global in nature, such as global warming that cuts across national boundaries and requires global cooperation in energy, environment, and economy to solve them. We are all in the same boat and must work together to meet these formidable challenges.

According to the International Energy Outlook 2011 reference scenario, the world’s energy consumption is expected to grow by 53% between 2008 and 2035. Global electricity generation will grow from 19.1 trillion kWh in 2008 to 35.2 trillion kWh in 2035, an increase of 84%. Likewise, nuclear generation is expected to increase from 2.6 trillion kWh in 2008 to 4.9 tril­lion kWh in 2035. As for Korea, energy is particularly crucial for its national growth planning, as Korea has virtually no natural resources.

Public concerns over the global ability to manage, and eventually dispose of RAW, especially HLW/SF, remain high. Emplacement in the deep geology is an internationally recognized disposal solution for HLW and SF, and China is planning to use this route. While China’s GDF programme is at an early stage, like all international waste management programmes imple­menting geological disposal, it is considering multi-barrier concepts com­prising engineered and natural barriers between the HLW/SF in the geosphere and the biosphere, while bentonite-based engineered barrier systems (EBS) were considered in China as early as the 1990s [16-17]. The current preliminary geological disposal concept for its HLW/SF is to use a shaft-tunnel model in the saturated zones of granite rock (Fig. 22.4). Over the past 20 years, China has made great strides in its geological repository programme including, as described above, geologically surveying the whole country for its georepository site selection and optimization of backfill/ buffer materials that will be needed for the GDF safety cases [18].

Many other countries are developing similar concepts for permanent disposal of radioactive waste deep underground: solidification of HLW/SF

Bentonite

backfill

A multi-barrier concept

Buffer

22.4 China s preliminary HLW repository concept.

using glass and ceramics, packaging in metal canisters, following temporary storage above ground before permanent geological disposal in natural barrier systems such as a granite rock-body, using a multi-barrier system [16,17]. Chinese researchers have suggested that EBS is a major component in guaranteeing long-term safety, making it necessary to conduct fundamen­tal research on the coupled THMC (thermal-hydrological-mechanical — chemical) behaviour of bentonite under simulated geological disposal conditions, and subsequently to reveal the property changes of the ben­tonite over a long period of time.

The requirements for HLW backfill materials are long-term chemical and physical stability, good mechanical properties, volume expandability in contact with water and very low water penetrability. Other requirements also include the ability to hinder nuclide migration, good thermal conduc­tivity and thermal stability, radiation resistance and stability, natural avail­ability and importantly, low cost.

Many years of research in Europe and China on bentonite backfill mate­rials for the EBS has revealed that bentonite comprising predominantly montmorillonite is considered to give the best performance in terms of low water penetration, high volume expansion, and excellent nuclide absorption and retention, as well as being abundant.

China is rich in mineral reserves and has large bentonite reserves suitable for the EBS backfill/buffer (at one site with a volume of 40 x 40 x 0.7 km) in China’s Inner Mongolia region near Beijing. Bentonite with high content of expandable montmorillonite has been found in an area named Gao — Miao-Zi (GMZ, which in English means Highland Temple). This single reserve, as shown in Fig. 22.5 is over 280 x 106 tonnes.

It is expected that the bentonite at Gao-Miao-Zi will be used in China’s HLW/SF geological repositories. This bentonite is being considered as a part of the EBS due to its ability to retain radionuclides and other hazardous materials. Prior to considering modular designs for canister encapsulation in the GDF, bentonite natural resources, raw mineral analysis, characteriza­tion and processing, need to be investigated, developed and optimized for large-scale cost-effective manufacture. To demonstrate the long-term safety of a GDF in China, the influence of the bentonite composition and the properties of the compacted block/brick must be studied.

Some large-scale mock-up facilities have also been built in China to test the efficacy of backfill/buffer materials such as bentonite with designed canisters. A China mock-up test was recently initiated after a long period of research conducted with international support. It is based on a prelimi­nary concept of the HLW granite rock environment repository in China [19] . It was developed to investigate the THMC properties of compacted GMZ-Na-bentonite as shown in Fig. 22.6 . which reveals the arrangement of compacted bentonite mineral blocks inside the mock-up test steel. The work has been carried out and led by the Beijing Research Institute of Uranium Geology (BRIUG) [20-22] . The device contains a heater, which

simulates the heat from a container of HLW/SF, placed inside the com­pacted GMZ Na-bentonite blocks with total dry density 1,600 kg/m3. Water inflow through the barrier from its outer surface simulates the water pen­etration. The device is a large steel tank of 900 mm internal diameter and 2200 mm in height. The experiment is being performed at the BRIUG labo­ratory and the design concept is shown in Fig. 22.6 and Plate VI (between pages 448 and 449). In Fig. 22.6, the compacted engineered bentonite blocks are arranged inside a mock-up test facility within a steel tank (top view). This mock-up THMC test consists of a heater (canister), bentonite blocks within a cylindrical steel tank, as shown in Plate VI as a sketch of the cross section of the China mock-up facility and the arrangement of central heater, steel canister, bentonite blocks/bricks and multiple sensor arrange­ment [22] .

About 100 out of the 2,200 locations used for soil sampling were used for the analysis of strontium-89 and 90. The locations where strontium-89, having a half-life of 50.5 days, was detected is attributable to the accident, while locations where only strontium-90 was detected should be attributed to the result of weapons test fall-out made before the accident. In fact, locations without strontium-89 showed strontium-90 concentration lower than 950 Bq/m2, which is the level of weapons testing fallout. According to the contamination map created by Japan’s Ministry of Education, Culture, Sports Science and Technology (MEXT), locations where both strontium-89 and 90 were detected are distributed over an area of about 50 km radius from the NPP. In the area within about 50 km radius, the maximum concen­tration of strontium-90 was 5,700 Bq/m2, but it corresponds to only 0.12 mSv dose over 50 years, which is much lower than the effect of cesium-134 and 137. The ratio between detected strontium-89 to 90 was in the range from 1.9 to 4.0, which is in the possible range of the measurement error to be associated with the difficult detection of в-rays. On the other hand, the ratio of strontium-90 to cesium-137 was found to vary extensively from 0.00016 to 0.058 depending on the location, which suggests a non-uniform distribu­tion of strontium compared with cesium. The ratio of radioactivity of strontium-90 to cesium-137 in the nuclear fuels in the reactor core is in the range from 0.7 to 1.0, therefore, the low ratio of strontium-90 observed for the deposition indicates the lower volatility or mobility of strontium than cesium, being the most volatile and movable element in the fuel components.

Strontium-90 of 195 Bq/kg was found in the sediment on the roof of an apartment building in the city of Yokohama, south of Tokyo, and this news caused a controversy about the possibility of long-distance transfer of stron­tium from Fukushima. However, it is understood to be attributable to the result of past weapons testing fall-out, and its attribution to the accident was incorrect.