The most important part of the DOE cleanup program is safety, which is integral to every program and project. In addition, DOE EM is implement­ing DOE Standard 1189 (DOE, 2008), which requires that safety-related documents and reviews be completed in the initial stages of the design process. DOE EM expects that integrating safety analyses up front in project design will avoid costly changes later (DOE, 2009).

Technology development is another key element of the cleanup program. The technology program is designed to provide a best-in-class science and engineering foundation and develop new technologies to reduce technical risk and uncertainty, support cleanup decisions, improve opera­tional efficiency, reduce costs, and accelerate schedules. In addition, laboratory — and pilot-scale testing is an important part of the technology maturation process.

The EM program has a strong commitment to reducing the technical risk of its programs and projects, and it is implementing two efforts to reduce those risks. This first is to conduct a Technology Readiness Assessment (TRA) to reduce the risks of deployment of a new technology. TRAs provide a snapshot in time of the maturity of technologies and their readi­ness for inclusion in the project. The results of a TRA assist program and project managers in developing plans to mature the technologies and to make decisions related to technology insertion. Eleven TRAs had been completed by the end of 2012.

The second effort is to conduct an External Technical Review (ETR) as one of several steps to ensure timely resolution of engineering and technol­ogy issues. The results of the reviews serve as a basis for developing strate­gies for reducing identified technical risks, and providing technical information needed to support critical project decisions. Twenty-five ETRs had been completed by the end of 2012.

Adhering to sound project management practices is essential. This includes, but is not limited to, developing comprehensive plans with a clear end-state for the site, defining clear project scopes, identifying and assessing risks, conducting system analyses, conducting peer reviews, establishing firm performance objectives, and anticipating unexpected outcomes.

The cleanup program would not be nearly so successful without the full involvement of its stakeholders, who provide insights and advice on how best to implement and improve it. The program has citizen advisory boards chartered under the Federal Advisory Committee Act at eight cleanup sites. The DOE also supports working groups with the National Governors Asso­ciation, National Conference of State Legislators, Energy Communities Alliance representing local governments, and State and Tribal Government Working Group. The DOE also works closely with its federal and state regulators to ensure that cleanup is being conducted in accordance with applicable laws, regulations, and compliance agreements, and in ways and according to schedules that protect public health and the environment (DOE, 2010). Continuous and transparent communication with stakehold­ers is vital.

The DOE’s cleanup mission poses unique, technically complex, and costly challenges, which can be achieved only through an exceptional workforce. The program’s 40,000 federal and contractor employees have the necessary skills and experience such that it is a world leader in the safe management and disposition of RAW and nuclear materials, as well as the remediation of contaminated facilities, soil, and groundwater (DOE, 2010).

In summary, the United States has extensive experience in cleanup of nuclear waste and facilities resulting from half a century of nuclear activities. The cleanup program has solved environmental problems that, at one time, seemed unsolvable; it will continue to make progress in solving the complex challenges it still faces (DOE, 2009).

EW contains such low concentrations of radionuclides that it does not require radiation protection provisions. EW may be re-used, disposed of in general waste landfill facilities, disposed of in hazardous chemical disposal facilities or be recycled. EW needs to be cleared/exempted in terms of approved clearance/exemption criteria [6].

Very short-lived waste (VSLW)

VSLW contains short-lived radionuclides with a longer lived radionuclide content within approved clearance/exemption criteria. VSLW is treated by decay to a point where the short-lived radionuclides have also reached approved clearance/exemption criteria, whereafter it is managed as EW.

Liquid RAW can be divided into process drains, floor drains and laundry drains based on the sources of waste generation. It is mainly generated from the clean-up and maintenance processes of reactor coolant and related systems containing radioactivity. In general, liquid RAW is treated with evaporators, demineralizers, and/or filters. The effluent is released to the sea after monitoring whether the radioactivity of liquid effluent is lower than regulatory limits. It is also common for liquid wastes to be treated with ultracentrifugation, ion exchange, and reverse osmosis.

The Ministry of Education, Science, and Technology (MEST Notice No. 2008-31) prescribes the effluent control limit (ECL) for liquid effluent being discharged into the environment at the restricted area boundary. Operators
must conduct periodic assessments for the expected off-site dose due to the liquid effluent discharged into the environment, and routinely report results to the regulatory body (Korea Institute of Nuclear Safety, KINS).

A large portion of Japan ’ s radioactive waste (about 50%) is stored in the radioactive waste management facilities at the nuclear facilities. About 1,690 canisters of vitrified product and about 380 m3 of liquid waste as HLW are stored in the reprocessing facilities at Tokai and Rokkasho (interim storage facility of the glass canisters), as of the end of March 2010. About

267.0 m3 of LLW (excluding used steam generators, spent control rods, disused channel boxes) are stored in all nuclear facilities in Japan as of the end of March 2008. Storage volume of LLW is made up of approximately

144.0 m3 NPP waste, approximately 25,000 m3 TRU waste, approximately

9.0 m3 uranium waste, approximately 65,000 m3 research waste, and approximately 24,000 m3 RI waste6.

The JNFL near-surface disposal facility with engineered barrier systems in place at Rokkasho, Aomori-Ken is in operation for LLW from commer­cial NPPs and about 219,000 200 L drums have been disposed of as of the end of March 2010. About 1,670 tons of very low level wastes resulting from dismantling of the Japan Power Demonstration Reactor (JPDR) were dis­posed of at the near-surface disposal facility without engineered barriers at Tokai. This disposal facility has been on hold since October 1997.

The International Atomic Energy Agency (IAEA) [28] and the World Health Organization [29] indicate that radiation exposure can result in both short-term and long-term effects in every body organ. In the Fukushima accident, public concern focused on both acute radiation sickness and increased long-term cancer risk [30].

Three months after the Fukushima accident, 20 teams were dispatched to Fukushima to screen for human radiation exposure. More than 5,000 people were screened by the staff of Hirosaki University (Fig. 24.4), and the results showed no acute radiation injuries [31]. The six deaths associated with the operation of the NPP were not attributable to exposure to ionizing radiation [ 11] . However, biological responses after exposure to radiation are time-dependent (Fig. 24.5), and the lack of symptoms appearing in the short term does not indicate freedom from long-term adverse effects due to radiation exposure. In the case of the Chernobyl accident, although no cases of cancer were confirmed to be due to the radiation exposure, some studies suggest that the risk of thyroid cancer for children living in nearby

areas could have been increased by a factor of 2 to 5 per 1 Gy of thyroid dose [30]. The Japanese government invited children exposed to the radio­active releases at different districts to be evaluated for thyroid disorders. As at December 31, 2011, 14,442 children had undergone screening, and no cases of fluid-filled cysts larger than 20 mm in diameter were found. Between January and March 2012, 27,467 more children were screened for thyroid. Furthermore, starting in April 2014, a follow-up thyroid screening will be performed on 360,000 potentially affected children once every two years until age 20 and every five years above the age of 20 [11]. No information is currently available concerning the progress of the studies planned for pregnant women and evacuees.

The method of thermal imagery is considered to be one of the most modern and effective methods of scanning terrestrial objects. For successful detec­tion and identification of the UGE-controlled objects, such imagery requires knowledge of the spectral characteristics of radiation, weakening of the pathway of the working range of wavelengths, as well as the characteristics and capabilities of equipment in the temperature and spatial resolution of a UGE. Efficiency of detection of thermal anomalies from space can be increased by multispectral imaging including the use of the visible spectrum that provides a higher quality of decoding images and binds heat-radiating objects to the terrain. Low-orbiting satellites or space stations may be used as carriers of the recording apparatus. Although satellites and space stations both have long orbital paths of observation, the long-term survival of thermal anomalies allows them to receive and store information on the same site area due to the lack of restrictions in the number of times they can review the UGE sites.

Transfer of infrared radiation on the ‘Earth-Space’ tracks took place in the spectral range from 8 to 14 microns (comparative assessments in some cases took into account the adjacent region of the spectrum). The radiation detector was focused on the thermal anomaly, with an ideal spectral char­acteristic in the range of wavelengths, located at the altitude of the space­craft orbit equal to 300 km. The zenith angle of sight ranged from 0° to 80°. The distributions of basic meteorological parameters are used to character­ize the atmospheric conditions in cloud-free atmosphere within their natural variability in the warm and cold periods of the year (McClatchey et al., 1972). Gas models include vertical profiles of pressure, temperature, density, and the amount of water vapor, carbon dioxide and ozone as meteorological parameters, to a greater degree of influence on the transfer of radiant energy in this spectral range. Aerosol atmospheric models include a set of basic types of aerosol particles (dust, water soluble, water-dust, soot parti­cles, acid aerosols, volcanic dust), the vertical distribution of their concen­trations, the spectral values of volume extinction a, the scattering в, and absorption 8 coefficients for local and continental aerosol types.

A quantity that must be determined is the extinction coefficient E (flux density of radiation from a source of unit power) as a function of orbital altitude H, the zenith angle of sight v, complex of meteorological parame­ters M, the spectral range ДА and calculated as a linear functional:

E(v, ДА) = TM (v, ДА) • TM (v, ДА) • Teax (v, M, ДА) • Tg (H, v), [27.1]

where TC is the attenuation due to the weakening of the molecular scattering of radiation, T^b is the weakening due to molecular (gas) absorp­tion of radiation, TO is the radiation attenuation due to scattering and absorption by aerosol and Tg is the radiation attenuation due to geometrical factors.

The function in Eq. [27.1] is calculated from the following relations:

Here z is the current height above the Earth, РДА is the transmission func­tion of the atmospheric gases, Ao = 0.55 micrometers, H is the ceiling of the
atmosphere equal to 80 km, and R is a distance from the source to the receiver.

The greatest difficulty in calculating the factors given in Eqs [27.2-27.5] is the calculation of the transmission functions in Eq. [27.3]. The methodol­ogy for calculating the transmission functions is chosen in accordance with the work of McClatchey et al. (1972) and allows determination of the attenuation due to a selective absorption of atmospheric gases and water vapor continuum absorption. Calculations of extinction coefficient are pre­sented in Fig. 27.7 as the E function of the zenith angle of sight v. The figure shows that the influence of aerosol extinction and molecular scattering is much weaker than the gas absorption. This explains the higher values of spectral transmittance in the cold season compared to the warm. From the graphs it follows also that a change in viewing angle from 0° to 80° for all weather conditions and satellite altitudes can be incorporated in a single change to the extinction coefficient E.

The energy flux density value of the extinction coefficient E should be multiplied by the flux of the intrinsic radiation source in the corresponding intervals. Self-radiation of the UGE thermal anomaly can be approximately estimated as gray-body radiation with a surface area equal to the square of light, which manifests itself in the thermal image. Assuming that the emis­sive capacity FAX(T) of the thermal anomaly is constant throughout the area, for typical sizes and temperatures, radiation flux density on orbit with a height of 300 km is in the spectral range 8-14 micrometers which is quite high for the infrared radiation quantity of about 10-9W cm-2.

It should be borne in mind that detection of thermal anomalies against the background of the outgoing radiation from the Earth and the atmos­phere depends on the response of the radiometer receiving element to temperature change, i. e. on the temperature contrast AT = T — Tbg or to change in the radiation flux density, i. e. on the energy contrast

27.7 Extinction coefficient depending on sight angle for high (a) and low (b) transparency of atmosphere: 1, AX = 4-5 micrometers; 2, 8-10 micrometers; 3, 10-12 micrometers.

K = (Ф — Ф6?)/(Ф + ФЬе). Here T and Tbg are the temperature of the thermal anomaly and background, and Ф and Ф^ are the relevant flux density in the orbit of the satellite/space station, provided that the spatial resolution of the detectors is close enough to the size of the thermal anomaly. Modern infrared receivers have a temperature coefficient of resistance, reaching tens of percent at 1°C. The typical thermal anomalies in the temperature contrast is AT = 10°C. Energy contrast reaches values of 0.07-0.09, if the pixel size does not exceed the size of the thermal anomaly, and decreases linearly with the increase of the former. Minimum resolvable contrast to existing energy equipment is 0.4% and corresponds to the range of varia­tion ratio of the characteristic size of thermal anomaly and a pixel in the range from 0.03 to 0.3, i. e. spatial resolution can substantially exceed the size of the thermal anomaly.

In addition to the spatial resolution, an important quantity for assessing the quality of the thermal and optical system is the probability of detecting thermal radiation from the object depending on signal/noise ratio, where the noise means temperature threshold of detection or, alternatively, tem­perature noise equivalent. The noise value is defined experimentally for specific optical systems and receives the equations of heat background light and varies from 0.1 to 0.2°C.

The method described provides a consistent and effective way if imple­menting operation related to decoding of information from the thermal anomaly as well as identifying its location and parameters. In this case, we consider the temperature field of the study area, mapped character images obtained under various shooting conditions, and analyze their dynamics with regard to the influence of all other factors. If the interpretation of thermal images is performed in conjunction with the data from visible or multispectral photography, it will facilitate recognition of terrain objects and allow the exclusion of anomalies of topographical nature, e. g. sun — warmed rock outcrops.

[1] Romuvaara (Kuhmo Municipality)

• Veitsivaara (Hyrynsalmi Municipality)

• Kivetty (Konginkangas Municipality)

• Syyry (Sievi Municipality)

• Olkiluoto (Eurajoki Municipality).

In 1992, TVO published a summary of the results from the site investiga­tions. After this phase in the siting process, Veitsivaara and Syyry were discarded, since they were considered less suitable than the remaining three sites.

[2] requirements for waste packages (e. g., surface dose rates, surface con­tamination, mass, leak tightness);

• requirements for waste forms (e. g., radionuclides content, composition and parameters of waste matrix and solidified waste, encapsulation material);

[3] 159 Member States (as of February 2013).

[4] In February 2013: Australia, Austria, Belgium, Canada, Czech Rep., Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxemburg, Mexico, Netherlands, Norway, Poland, Portugal, Rep. of Korea, Russian Federation, Slovak Rep., Slovenia, Spain, Sweden, Switzerland, Turkey, United Kingdom and the United States of America.

[5] The European Commission (EC) establishes policies, directives, regula­tions and recommendations in the field of nuclear energy, including the safe management of spent fuel and radioactive waste. In 2007, following a decision of the EC, the European Nuclear Safety Regulators Group (ENSREG) was established as an independent, authoritative expert body. Its aim is to help to establish the conditions for continuous improvement and to reach a common understanding in the areas of nuclear safety and radioactive waste management. It is composed of senior officials from the national nuclear safety, RAW safety or radiation protection regulatory authorities from all 27 Member States in the European Union and representatives of the EC [22]. Recently the European Commission approved a new Directive on the management of spent fuel and radioactive waste [23].

• Regulatory associations (networks) in Africa (Forum of Regulatory Bodies in Africa — FNRBA [24]), Europe (WENRA — Western Euro­pean Nuclear Regulators Association [25]), Latin America (Latin American Forum of Nuclear and Radiological Regulatory Organisa­tions — FORO [26]), Arab countries — ANNuR (Arab Network for Nuclear Regulators [27]), Asia — Asian Nuclear Safety Network (ANSN

[6]The standard is entitled ‘Radiation Protection and Safety of Radiation Sources — Interna­tional Basic Safety Standards’, but is commonly referred to as the Basic Safety Standards or

BSS.

[7] Category 1 sources, if not safely managed or securely protected, would be likely to cause permanent injury to a person who handled them, or were otherwise in contact with them, for more than a few minutes. It would probably be fatal to be close to this amount of unshielded material for a period of a few minutes to an hour. These sources are typically used in practices such as radiothermal generators, irradiators and radiation teletherapy [50].

[8] Category 2 sources, if not safely managed or securely protected, could cause permanent injury to a person who handled them, or were otherwise in contact with them, for a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioac­tive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachy — therapy [50].

[9] Classification of Radioactive Waste, No. GSG-1 [3] that substitutes the previous waste classification No. 111-G-1.1 of 1994 [64];

• Management of Low and Intermediate Level Waste, No. WS-G-2.5 [65];

• Management of High Level Waste, No. WS-G-2.6 [66] ;

[10] 1 Tera-Becquerel (TBq) = 1012 atoms decaying per second or transmutations per second.

*1996-June 2012.

[11]1996-2002 — mission complete.

* 1991 to April 2012 at 142 L glass per canister and an assumed glass density of 2.75g/cc (390kg glass per container).

** Maximum total is 10,000 (capacity of vitrified product store), of which -2,200 will be returned to overseas customers. Actual total is expected to be less depending on post-operation clean-out strategy.

‘ 1989-2011.

11978-2008.

51985-1991.

"1995-2012.

“acidic waste loadings comprise fission products and minor actinides; corrosion products and alkali are not included as for neutralized wastes. a From [163]. bFrom [204]. cFrom [205].

dCaterine Veyer of AREVA, personal communication (2012). sSeiichiro Mitsui of JAEA, personal communication (2010). fP. P. Poluektor, personal communication (2010).

[12] single-phase (homogeneous) glasses

• multi-phase glass composite materials (GCMs; heterogeneous glasses)

• single-phase crystalline ceramic/mineral analogs

• multi-phase crystalline ceramic/mineral assemblages

• bitumen

• metals

• cements

• geopolymers (inorganic) and organic polymers

• hydroceramics

• ceramicretes.

[13]SRO: radius of influence -1.6-3 A around a central atom, e. g. polyhedra such as tetrahedral and octahedral structural units.

[14] MRO: radius of influence -3-6 A encompasses second — and third-neighbor environments around a central atom. The more highly ordered regions, referred to as clusters or quasicrystals, often have atomic arrangements that approach those of crystals.

[15] LRO extends beyond third-neighbor environments and gives crystalline ceramic/mineral structures their crystallographic periodicity.

[16] Supercalcines were the high temperature silicate-based ‘natural mineral’ assemblages proposed for HLW waste stabilization in the United States (1973-1985).

Adapted from [11].

[17] Phosphate glasses (aluminophosphates and iron phosphates) are not used commercially as frequently as the borosilicates and hence are not as well studied in HLW stabilization applications.

[18]This is not the baseline AJHM process that will produce a homogeneous glass with minimal crystallization.

substituted phase

These types of crystal-chemical substitutions have been studied in (1) Synroc (Synthetic rock) titanate phases such as zirconolite (CaZrTi2O7), perovskite (CaTiO3), and hollandites (nominally Ba(Al, Ti)2Ti6O16) [90], and (2) in high alumina tailored ceramic phases such as magnetoplumbites (Table 6.8). The magnetoplumbites (discussed below) are also found as a minor component in Synroc when the waste being stabilized is high in Al [91].

In the Synroc phase assemblages, the hollandite phase is the Cs+ host phase. The structure can be written as BaxCsy(Al, Fe)2x+yTi8-2x-yO16 where x + y must be <2 [92] [ There are two types of octahedral sites. One accommo­dates trivalent cations like Ah+, TrA and FeA while the other accommo­dates Ti4+ . The Cs+ is accommodated in tunnels that normally accommodate the Ba[+ cation. The Cs-Ba lattice sites are VIII-fold coordinated [90, 92] [ 7 Note that the number of lattice sites have to be equivalent on the left-hand side and right — hand sides of the equation.

[20] Saltstone contains 5 wt% cement, 25 wt% flyash, 25 wt% blast furnace slag, and 45 wt% salt solution.

[21] Conditions affecting dissolution, solubility of actinides. Environmental conditions such as reducing aqueous groundwater result in very low dissolution rates of fission products and low solubility of actinides in SNF dissolution. This limits the radionuclide release into the biosphere.

[22] —

[23] evaluation of the severity of the problem in terms of radionuclide con­centration or dose levels to determine whether there is a need to remediate;

• evaluation of the remediation alternatives including the feasibility, cost, waste generation and management, and risk reduction;

[24] De-licensing is taken to mean ‘ending of the period of responsibility under the Nuclear Installations Act’ and happens when the HSE gives notice in writing to the operator that in its opinion there has ‘ceased to be any danger from ionizing radiations’.

• Any residual radioactivity, above natural background levels, which can be satisfactorily demonstrated to pose a risk less than one in a million per year (of the order of 10 microsieverts or less per year) for any reasonably foreseeable land use is taken to be broadly acceptable.

• Additionally, the operator should demonstrate that risk has been reduced to levels as low as reasonably achievable and should take into account the views of relevant regulators in respect of non-radiological contamination issues.

• All risks are taken to be additional to natural background levels for the area, including an allowance for impacts from authorized discharges and artificial background from worldwide sources.

• The IAEA safety guide on the application of the concept of exclusion, exemption and clearance (RS-G-1.7) (IAEA, 2004) contains radionu­clide specific values that should be used to demonstrate achievement of

[25] repeated grouting using highly penetrating biocide cement composi­tions of cemented solidified RAW;

• creation of an anti-filtration screen in the soil on the perimeter of the repository;

• formation of a barrow from the natural materials on the surface of the repository;

• piling and welding sheets of the geo-membrane Carbofol (1.0-2.0 mm thick);

The main characteristics of the ‘Shelter’ object (SO) RAW are given in Table 11.7. The total waste activity of the SO as of the beginning of 2005 is approximately 4.1 x 1017Bq, and the waste volume (according to different estimates) is between 530,000 and 1,730,000 m3.

The volume of waste concentrated in RWDP and the main RWTSP of the ChEZ is approximately 2 million m3, and the total activity is estimated at 7.7 x 1015Bq. It should also be noted that the same amount of radionu­clides is again contained in natural objects (vegetation, soils, bottom

[27] In Ukraine, activities to create a deep geological repository have been carried out since 1993. They are performed by Institutes of the National Academy of Sciences of Ukraine and the enterprises of the State Geological Survey. This activity refers to the early stages of a siting and conceptual repository design. It is assumed that the most promising host rocks in which to locate a geological repository are Archaean and Proterozoic crystalline rocks of the Cher­nobyl Exclusion Zone and its vicinity. Two possible options for the repository design are considered: a mine (KBS-3 concept, Sweden) and a borehole one (VDH concept, Sweden). Further information can be found in Shestopalov et al. (2005,2008).

[28]The London Convention subdivided radioactive waste into high and low level waste, with definitions of high and low level waste that were derived specifically for disposals at sea.

[29] The policy is enabling to allow waste managers, regulators, facility owners and the NDA to take decisions on the long-term management of HAW.

• The policy is not prescriptive and it is the responsibility of HAW manag­ers to decide on HAW management methods on a case-by-case basis in accordance with the policy framework.

• An implementation strategy for the policy will be developed by the Scottish government.

• Long-term storage is the primary long-term management option.

• The waste hierarchy should be applied and HAW can be managed by treatment, storage or disposal.

[30] Defense waste is mainly characterized as radioactive material in a very diluted form, whereas civilian waste is mainly generated in a concen­trated form.

[31] CANDU® is a registered trademark of Atomic Energy of Canada Limited.

[32]Till or glacial till is unsorted glacial sediment. Glacial drift is a general term for the coarsely graded and extremely heterogeneous sediments of glacial origin. Glacial till is that part of glacial drift which was deposited directly by the glacier. Its content may vary from clays to mixtures of clay, sand, gravel and boulders.

[33] Although no medium depth (higher confinement) repositories cur­rently exist in South Africa (still in the planning phase), waste needs to be classed in terms of general criteria for effective pre-disposal man­agement. Waste characterized in terms of general criteria will be con­sidered in the long-term safety assessments that are necessary for the authorization of such repositories. Taking a retrospective approach, the design of repositories will have to be suitable for waste that has been processed and is in compliance with specific long-term safety-related criteria.

• The long-term safety of the national near-surface repository for LILW at Vaalputs in the Northern Cape is demonstrated and is currently authorized in terms of specific criteria. The long-term safety assessment of Vaalputs needs to be reviewed in terms of specific criteria prior to the authorization of receipt of waste from different generators. This is necessary to evaluate the suitability of the disposal system at Vaalputs for specific waste streams and additional inventories.

[34] waste treatment and volume reduction technology

• low-level waste vitrification technology

[35] safety concerns about the disposal facility,

• lack of transparency and fairness during project implementation,

• lack of social consensus among the stakeholders.

In February 2004, the Ministry of Knowledge Economy (MKE) announced new site selection procedures, and MKE/KHNP made various efforts to enhance the acceptance by local residents of disposal facilities. As a result,

[36] groundwater infiltration rate into silos: re-estimation of the groundwa­ter infiltration rate into the concrete

• silos during the post-closure phase, in combination with justification of the human intrusion scenarios

• quality control of geochemical data: reconfirmation of the representa­tiveness of empirically determined site-specific geochemical data (e. g. sorption coefficients, diffusion coefficients, etc.)

• long-term management of uncertainties in geochemical data

• seismic safety and design: verification of the geological structure model and tectonic activity of the site

• structural stability of the rock caverns and silos.

The above KTIs were resolved through regulatory dialogues and requests for more detailed information along with the applicant ’s amendments to the license application documents, reflecting the results of further supple­mentary site surveys, safety assessments, and design changes, which occurred during the review process.

[37] control of waste transfers to prevent contamination,

• maintenance of normal operations of the waste treatment system to reduce generation of secondary waste,

• minimization of the entry of materials into controlled areas, and

[38] replacing dismantling notification by licensee, to approval of the licen­see’s decommissioning plan by the regulatory body,

• implementation of decommissioning as approved in the decommission­ing plan,

• completion of decommissioning is confirmed by the regulatory body and after confirmation of the completion of decommissioning, the operating licence becomes ineffective,

• the regulatory activities during the decommissioning process should be changed in accordance with the changes of functions of facilities and safety operation activities as the decommissioning proceeds.

Source: Used with permission of the Ministry of Economy, Trade and Industry (METI).

maximum at the end of March, and it gradually decreased to 100 Bq/L in May.

The release of volatile radioactive nuclides into the atmosphere from the three units is considered to have occurred mainly after March 14, while the hydrogen explosions of units 1 and 3 occurred on March 12 and on the morning of March 14, respectively. These large releases after the night of March 14, along with the unfortunate climate conditions of wind and rain/ snow at that time, have probably caused contamination over a wide region of the Fukushima Prefecture in a north-easterly direction. Along with the varying climate conditions, particularly of wind direction, some of the

The first task assigned to ANDRA was to operate the surface disposal of SL-LILW that had been created in 1969 at the Centre de la Manche. It also laid down some rules to secure and streamline disposal of waste. For example, the waste had to be packaged in standard packages. In addition,

ANDRA built a collection system to monitor and control the water coming out of the disposal facility, which allowed the impact of the centre on its environment to be monitored.

A new disposal centre in the Aube district

From 1984, ANDRA began looking for a new site for a disposal facility to replace the Centre de la Manche. Geological studies were undertaken in different ‘Departments’ (the ‘Department’ is the main political and admin­istrative subdivision in France). In 1984 and 1985, more than 500 boreholes were drilled in the Aube Department to select a specific location. At this time, ANDRA perfected the technique of using a multi-barrier system consisting of the package, the engineered barrier and the geology to dispose of the waste. Meanwhile, local consultation was carried out through the organization of several visits and meetings with local stakeholders. On 22 July 1987, the Prime Minister signed the Declaration of public interest: the new disposal facility for SL-LILW, the CSFMA, was located in the Aube Department, near the village of Soulaines-Dhuys.

LLW disposal in the UK has been ongoing since the 1950s, providing con­siderable perspective on the approaches to, and the strategy development of, LLW disposal. Since 1959, LLW has been disposed of at the UK’s national low level waste repository (LLWR) in Cumbria, in addition to a number of other LLW disposal sites including various Sellafield pits, Hunterston A and Dounreay. LLW strategy development has largely progressed in response to the need to preserve the capacity of the LLWR for as long as possible given future arisings of LLW in the UK.

The LLWR (Fig. 16.4), located on the site of a second World War muni­tions factory, was initially developed as a series of excavated trenches into which wastes were loose tipped between 1959 and 1995 (LLWR, 2011). The trenches were designed with drainage and runoff collection systems and were largely keyed into a low hydraulic conductivity clay layer. Where this was absent, bentonite was rotovated into the trench base.

In 1988, trench disposals were phased out in favour of LLW disposal to engineered vaults. A number of improvements were made to the trench disposal areas after this time. The installation of an interim cap over Trenches 1-6 took place in 1989 to minimise rainfall ingress into the wastes and a bentonite cut-off wall was excavated on the north and east sides of the disposal area in 1988 to reduce the potential for tritium migration in ground­water (LLWR, 2011 ).

Overall, the major components of LLW are building rubble, soil and steel items from the dismantling and demolition of nuclear reactors and other nuclear facilities (Fig. 16.7).

LLLE is a continuous product from current decommissioning and current waste management operations. The LLLE is discharged to sea through a system that was refurbished in 1992 and a LLLE treatment plant (LLLETP) that was brought into operation in 1997 to replace the original one. All discharges have been, and are, authorised by SEPA.

ILW

ILW has been produced from maintenance, waste management and reproc­essing activities associated with the fast reactor research programme. ILW is currently being produced from decommissioning of the facilities. As the fast reactor fuels contained plutonium, there is significant alpha contamina­tion associated with many of Dounreay ILW streams. This leads to specific handling and containment requirements in all waste management and decommissioning operations.

SNF results from the once-through fuel cycle (i. e., no further processing conducted). It contains greater than 99% of the radioactivity and has unique characteristics compared to wastes from fossil plants. Because only about 5% of the energy value has been consumed in the reactor, it can also rep­resent a future energy resource. The energy release from nuclear fission per ton of fuel is about a million times greater than the energy release from the burning of fossil fuels. The waste volume generated is about a million times less. The quantity of SNF is small per unit of energy produced. The small quantity (-20 tons per reactor per year) makes multiple waste management options economically feasible: multiple direct disposal options and multiple options to process the SNF chemically for recovery of selected materials for recycle and/or conversion into different waste forms.

Reactors discharge SNF that contains fissile materials (fuel) and fission products (waste). The radioactivity and decay heat of SNF decreases rapidly with time; thus, to reduce handling risks and costs, SNF is stored before transport, disposal, or recycling. SNF storage is a required step in all open and closed fuel cycles. This is a consequence of the nuclear characteristics of SNF. The radioactivity decreases rapidly with time, resulting in radioac­tive decay heat and gamma radiation decreasing rapidly with time. There are large safety and economic incentives to allow the radioactivity of SNF to decrease before transport, processing, or disposal.

Upon reactor shutdown, SNF is intensely radioactive and generates large quantities of decay heat — equal to about 6% of the power output of the reactor. However, the radioactive decay heat decreases very rapidly reach­ing 0.5% in one week. The refueling strategy in light water reactors (LWRs) is to transfer the SNF from the reactor core to the SNF storage pool where the water provides cooling and radiation shielding. After about ten years, the radioactivity will decrease by another factor of 100.

If SNF is to be disposed of in a repository, it will likely be stored for approximately 40-60 years prior to disposal. Peak temperatures in a geo­logical repository are limited to ensure long-term repository performance. If the temperatures are too high, the performance of the waste form, waste package, and geology may be impaired. Peak repository temperatures would be controlled by limiting the allowable decay heat per waste package. If the SNF is stored for several decades, several advantages would result: the decay heat per ton of SNF decreases; more SNF can be placed in each waste package; the waste packages can be spaced closer to each other underground; the size (footprint) of the repository is reduced; and the cost of the repository is reduced. Like SNF, the HLW will be cooled for 40-60 years before ultimate disposal to reduce the decay heat.