The RAW classification scheme as applicable to Necsa is generally in com­pliance with the scheme as published in the NRWMPS [3] . The proposed scheme for Necsa also makes provision for the latest IAEA international developments in RAW classification [4]. More waste classes are covered to address all the various waste streams of Necsa (e. g., special waste). The additional waste classes could be regarded as sub-classes of the main classes [3] which facilitate effective waste management and are complementary to the management methodology. The latest waste classification guidelines [4] are based on long-term safety aspects and do not specify criteria to distin­guish between the classes. For the Necsa classification scheme, some criteria are maintained in view of the following: [33]

The Necsa solid radioactive waste classification system is presented sche­matically in Fig. 20.6. The classification scheme also complies with the general classification scheme as indicated in Table 20.1.

Spent fuel (SF) generated from nuclear power plants has been stored in spent fuel storage pools at reactors or in on-site dry storage facilities. Dry storage is currently used only for PHWR (CANDU) spent fuel sufficiently decayed for about six years in storage pools. The low — and intermediate — level radioactive waste (LILW) generated from the NPPs has been stored in on-site radioactive waste storage facilities.

Radioactive waste materials are also generated from fuel fabrication processes and they are stored on-site. In addition, the use of radioactive materials in medicine, research work and industry has increased steadily. These facilities are located throughout the country and generate various types of RAW. Radioisotope (RI) contaminated waste from these facilities is stored at an RI waste management facility. There has been much turmoil concerning public acceptance issues associated with the LILW disposal facility site selection, with a number of unsuccessful attempts to select the site.

The Korean government has striven to secure a disposal site for the safe management of RAW since the early 1980s. After a number of failed attempts, the Korean government issued a Public Notice on the selection of a candidate site for the LILW disposal facility, and the city of Gyeongju was selected as the final candidate site in November 2005 following the procedures involving a site suitability assessment, local referenda, etc. as specified in the Public Notice. The Korea Radioactive Waste Management Corporation. (KRMC) was established in 2009 as a new Korean RAW management agency and is currently undertaking the construction of the LILW disposal facility in accordance with the permit issued.

Spent fuel generated from NPPs is stored in the spent fuel storage facility in each unit. The storage capacity for spent fuel has been expanded as a consequence of the delayed construction schedule of the away-from-reactor (AFR) interim storage.

The Law of the People’s Republic of China on Environmental Impact Assessment states that for projects that may have adverse environmental impact, public meetings should be held or other approaches adopted to solicit comments on the draft environmental impact assessment statement from relevant organizations, experts and the public before its submission for review and approval. The constructor and the operator of the proposed site will need to take into consideration the comments provided from the relevant organizations, experts and the public, and provide additional explanations on whether these comments have been incorporated when submitting the environmental impact assessment report for review.

The spread of radioactive pollutants was not confined to Japan. Due to the prevailing westerlies during the accident, the radioactive nuclides had the potential to be transported offshore, across the Pacific Ocean, and further to the North American continent. Monitoring of seawater, soil and atmos­phere was being done at 25 locations on the plant site, 12 locations on the boundary, and other locations further afield [1]. Trace amounts of radiation, including iodine-131, cesium-134 and cesium-137, were being observed around the world (New York State, Alaska, Hawaii, Oregon, California, Montreal, and Austria) [13]. Radioactive isotopes originating from Fukushima were picked up by over 40 CTBTO radionuclide monitoring stations [15] .

On March 17-18, 2011, the first arrival of the airborne fission products, iodine-131, iodine-132, tellurium-132, cesium-134, and cesium-137, was detected in Seattle, Washington (USA) by their characteristic gamma rays. Leon et al. [16] used a NOAA HYSPLIT model to assess their transport time and possible trajectories across the Pacific. Plate XII (between pages 448 and 449) shows three trajectories of the radioactive nuclides, which indicate the range of transport pathways. The start time was set to March 12, 2011 at 10 UTC (Coordinated Universal Time), which was approxi­mately 3 hours after the reported explosion from unit 1. The trajectories were calculated for three heights in the atmospheric boundary layer: 500, 1,000, and 1,500 m above ground level. The 500 m trajectory was found to be caught up in and raised by a cyclonic system over the Bering Sea. The trajectories started at 1,000 m and 1,500 m were also partially lofted but did not get involved in the cyclonic pattern. Instead, they were found to be rapidly transported across the Pacific. Upon arrival at the west coast of the US, the transport again split, with one arm transported to the north in a cyclonic direction around a low pressure system located off the coast of Washington state. There were rain showers and cool weather in western Washington at the arrival time of the plume, and the strong divergence and precipitation associated with these weather systems most likely significantly reduced the concentrations of radionuclides that were transported. The trajectory initially started at 1,500 m was transported in the boundary layer towards California. Overall, the trajectories support the notion of transport of the radionuclides from the Japanese boundary layer to the US boundary layer in only 5-6 days. This result is significantly faster than the other previ­ous work which examined the trans-Pacific transport times [17], especially considering the radionuclides were released in the boundary layer over Japan and measured in the boundary layer along the US west coast.

After crossing the North American continent, the contaminated air masses were anticipated to continue to move towards the North Atlantic and reach Europe. The first sign of the radioactive material in Europe was detected on March 19, 2011 in Iceland, 7 days after the explosion of the unit 1 reactor. On March 23-24, most European countries had detected the radiation. Around March 28-30, the first radioactive peak was observed. The time — and spatially-averaged values from March 20 to April 13 in Europe were 0.076 and 0.072 mBq/m3 for cesium-137 and cesium-134, respectively [ 18]. Cesium-137 airborne activity levels reported after the Fukushima Daiichi NPP incident were at least 10,000 to 100,000 times lower than those observed after the Chernobyl accident. Regardless of the radio­nuclide considered, airborne activity levels remained sufficiently low as to be of no concern to public health in Europe. Although the prevailing wind during the accident was westerly, the radiation effect in Hong Kong, more than 2,000 miles southeast from Fukushima, was also detected before April 14, 2011. An activity of iodine-131 of 62.5 pBq/m3 was first detected on March 26, 2011, and a maximum value of 828 pBq/m3 was observed on March 29, 2011, in Hong Kong [19].

These final sections overview the results of flow and transport modeling studies and assessments of uncertainty for the Frenchman Flat CAU, the most developed of the CAU studies on the NNSS. There are three dominant features of all conceptual models of the Frenchman Flat basin (Fig. 26.4):

1. the high hydraulic heads in the CP basin northwest of Frenchman Flat (over 100 m higher heads than the Frenchman Flat basin; see Fig. 26.4),

2. the semi-perched condition of groundwater in the alluvial and volcanic aquifers with higher heads in these aquifers than the regional LCA,

3. the southeastward thinning of the volcanic section beneath the basin across Frenchman Flat.

These combined features support two inferential observations for the basin. First, groundwater flow in the alluvial and volcanic aquifers is likely hori­zontal across the basin from northwest to southeast (NNES, 2010a, b). Second, there is increased leakage downward into the LCA from the allu­vial and volcanic aquifers as the basal volcanic confining unit thins to the southeast and/or is offset by faults associated with the Rock Valley fault system. Particle track studies originating at locations of underground tests show southeast flow through the alluvial and volcanic aquifers changing to southwestward flow in the LCA following surface and subsurface faults associated with the basin structure (Bechtel Nevada, 2005; SNJV, 2006; NNES, 2010a, b). These observations are consistent with groundwater flow converging into and following faults of the Rock Valley fault system in southern Frenchman Flat (Fig. 26.8).

Modeling studies for the Frenchman Flat CAU combine steady state and transient source term studies, multiple alternative representations of the groundwater flow system, and probabilistic transport simulations. Source term models of radionuclide releases into groundwater were developed for

26.8 Satellite photograph of the Frenchman Flat basin on the southeast edge of the NNSS showing the major structural features of the basin and directions of groundwater flow (large black arrows: regional flow system; large gray arrow: local flow in the alluvial and volcanic aquifers). The Rock Valley fault zone is a zone of echelon faults that form the Rock Valley fault system. The asterisks mark the location of ten underground nuclear tests; three in central Frenchman Flat and seven in the north part of the basin. The solid gray lines outline the edges of contaminated groundwater defined by the 95th percentile of exceeding the radiological standards of the Safe Drinking Water Act over 1,000 years. These contaminant boundaries are small (<500 m length and for some tests in alluvium, the contaminant boundaries are smaller than the asterisk symbol marking the test locations); the contaminant boundaries are larger for two tests where the underground cavity was in or near fractured volcanic rocks (two tests in the northern area) or where a 17-year radionuclide pumping experiment discharged contaminated groundwater on the surface (one test in the central area).

two settings. First, the radiological source term for underground tests in alluvium were calibrated, for both steady-state and transient models, to observed breakthrough of radionuclides at a pumping well located 91 m from the CAMBRIC test in the water table in alluvium (Tompson et al., 1999; Carle et al., 2007). Second, two underground tests in northern French­man Flat were conducted above the water table in or near fractured vol­canic rock, where the rock permeability and porosity is inferred to be enhanced from the effects of the test detonation (IAEA, 1998). Simplified source term models were developed for these tests that account for unsatu­rated and saturated flow and transport and test-induced changes in rock properties (NNES, 2010a, b).

Multiple steady state groundwater flow models were developed for the Frenchman Flat CAU (SNJV, 2006) that are calibrated to hydraulic heads and permeability data for hydrostratigraphic rock units, and attempt to account for conceptual model uncertainty. The evaluated components of conceptual (structural) model uncertainty include variability in boundary conditions and boundary fluxes, permissible alternative hydrogeological frameworks for the basin, including structure (faults and basin features), stratigraphic units within the basin, and alternative recharge models. The goal in developing flow models was not to identify a best-fit calibration or a best predictor flow model but instead to distinguish a range of alternative flow models that capture the range of variation in flow fields from paramet­ric and structural uncertainty. This range in groundwater flow was then used in transport simulations. Statistical metrics of goodness of fit of alternative groundwater calibrations did not provide useful information for discrimi­nating or screening groundwater flow models. Two alternative sets of data did provide useful information for categorizing results for calibrated flow models (SNJV, 2006). These include variability in particle track results, and variability in groundwater velocity and direction at test cavity locations using linear predictive uncertainty analysis from parameter estimation soft­ware (PEST; Doherty, 2007).

Monte Carlo transport simulations were conducted for underground tests at the two testing areas in Frenchman Flat (Fig. 26.4). Four flow models were combined with alternative sets of boundary conditions (boundary fluxes, hydrostratigraphic frameworks and recharge) to represent the vari­ability in the groundwater flow field (velocity and direction of flow at the test cavity). These flow conditions were established at the underground test cavities as the initial conditions for transport simulations sampling stochas­tic transport parameters using a streamline-based convolution transport code (Robinson et al., 2011). Radionuclide concentrations for 1,000 years of transport were post-processed to develop probabilistic forecasts of exceeding the radiological requirements of the SDWA (Fig. 26.8); the boundary of this representation denotes the limits of contaminated groundwater (contaminant boundary) defined as a 5% chance or less of exceeding the SDWA. There are two categories of contaminant boundaries: (1) small boundaries (<500 m maximum lateral distance) where the test cavity and transport are in the alluvial aquifer and (2) larger boundaries (>1600 m) where the source term and/or transport is in fractured volcanic rock. For the latter category (two underground tests), the contaminant boundaries extend slightly off the NNSS boundaries into adjacent Federal land (Fig. 26.8).

The contaminant boundaries of the central testing area of Frenchman Flat (Fig. 26.8) are complicated by two factors. First, the long-term pumping test for the CAMBRIC test discharged contaminated groundwater on the surface into a ditch that drained into the Frenchman Flat playa. Second, the discharged contaminated water in the drainage ditch infiltrated to the water table in concentrations that exceed the SDWA. This required transient models to account for the 17 years of continuous aquifer pumping and surface discharge of contaminated water and significantly extended the contaminant boundaries of the central testing area.

The contaminant boundaries depicted in Fig. 26.8 will be used for two regulatory decisions. First, the boundary geometries will be used to desig­nate surface use restriction areas where institutional controls will be imposed to restrict all drilling to potentially contaminated groundwater. Second, the contaminant boundaries and results of subsequent monitoring studies will be used by NDEP to identify a regulatory boundary designed to protect the public and environment from exposure to contaminated groundwater. The NNSA/NSO will be required to develop a plan to miti­gate potential impacts on the public, if radionuclides are detected at the regulatory boundary. The regulatory boundary has tentatively been identi­fied as the Rock Valley fault zone at the southern end of Frenchman Flat, the expected migration pathway to public access to groundwater south of the southern boundary of the NNSS.

The transport model for the Frenchman Flat CAU was accepted by NDEP following successful external peer review of the CAU studies (Navarro-Intera, 2010). This marks the first successful completion of the model development stage under the UGTA strategy and the initiation of the model evaluation stage for the Frenchman Flat CAU (USDOE, 2011).

26.4 Acknowledgments

This chapter is a snapshot of ongoing work for UGTA. This work is a multi­disciplinary cooperative effort conducted by scientists at the Desert Research Institute, the Lawrence Livermore National Laboratory, the Los Alamos National Laboratory, National Security Technologies, Navarro — Intera and the US Geological Survey. The chapter was improved through review comments provided by Bimal Mukhopadhyay, Joe Fenelon and Susan Krenzien. Nathan Bryant and Joe Fenelon assisted in the develop­ment of the chapter figures.

The Gorleben facility is located in an undisturbed salt dome near the village of Gorleben approximately 100 km southeast of Hamburg, Germany. Fol­lowing the selection of the Gorleben site in 1977 for investigation as a poten­tial repository for heat-generating wastes, and the establishment of DBE in 1979, a comprehensive surface-based investigative programme was initiated to characterize the salt dome and the surrounding area of the site. Based on the positive indications from the surface investigations, an underground exploratory facility was designed and constructed by DBE in 1986 on behalf of the BfS. The Gorleben exploratory facility was intentionally designed to facilitate conversion to a repository, assuming subsequent investigations would continue to support the site’s suitability. From 2000 to 2010, site char­acterization activities at Gorleben were suspended by the federal govern­ment as part of a moratorium agreement negotiated between the previous government and the nuclear industry. In October 2010 the moratorium expired and site characterization and licensing activities were restarted.

In the 1980s and 1990s, considerable effort was invested in investigating the Gorleben salt dome as a potential site for hosting a nuclear waste repository. The investigations supported the concept of rock salt as a host environment based on its very low inherent permeability and the self­healing nature of fractures due to the plastic response behaviour of the rock type. In addition to the subsurface research facility, many of the surface installations were also completed prior to the imposition of the ten-year moratorium. In the framework of research, development and demonstra­tion activities, significant advances have been made with respect to proto­type equipment development, including development of a shaft hoist system with a capacity to lift 85 tonnes, emplacement machines for both drift and borehole disposal, and equipment for backfilling disposal drifts. For these reasons, the facility at Gorleben is unique when compared to other inter­national repositories in that much of the site characterization and surface infrastructure work was actually completed in the 1980s and 1990s. As a result, despite the moratorium, Gorleben remains one of the most techni­cally advanced potential high-level RAW repository sites currently under consideration both in a national and international sense. Figure 14.7 shows the potential Gorleben repository concept and existing prototype equipment.

Shaft transport

Borehole emplacement

Backfilling slinger truck in a disposal drift

Drift emplacement

14.7 Gorleben repository concept with prototype shaft hoist, borehole emplacement machine, backfilling slinger truck and drift emplacement machine. Source: Provided by the German Company for the Construction and Operation of Waste Repositories (DBE), Peine, Germany.

Since the moratorium was lifted and research was recommenced, new safety requirements for the disposal of heat-generating waste, as well as requirements for retrievability have been published and are expected to be enacted. The performance criteria include the evaluation of repository safety for a one million-year period (referred to as the period of geological stability) at an annual effective exposure not to exceed 10 pSv for likely event scenarios and 100 pSv for less likely events (BMU, 2010). However, recent legal actions challenging key aspects of the operating licence for the Gorleben site investigation, submitted to the Upper Administrative Court of Luneburg, have resulted in the suspension of on-going subsurface research activities at Gorleben with immediate effect pending further judi­cial review.

A preliminary safety assessment (vorlaufige Sicherheitsanalyse fur den Standort Gorleben, VSG) that will provide a detailed evaluation of the potential suitability of the Gorleben salt dome as a repository host formation for the disposal of heat generating waste is currently being com­pleted. The Gesellschaft fur Anlagen — und Reaktorsicherheit (GRS) is responsible for developing the VSG in collaboration with a team of con­tributing organizations, and is scheduled to be completed in 2013.

Applications for approval of RAW disposal facilities made to the environ­ment agencies and planning authorities under EPR10 (as amended) must be supported by an environmental safety case. Requirements for authorisation are set out in a guidance document published jointly by the environment agencies, ‘Near-surface Disposal Facilities on Land for Solid Radioactive Wastes: Guidance on the Requirements for Authorisation’ (the GRA). This was originally published in 1997 (Environment Agency et al., 1997) and revised and updated in 2009 (Environment Agency et al., 2009). An equivalent document sets out the guidance for disposal of geological waste, although the guidance for deep geological disposal does not apply in Scotland. (Environment Agency and Northern Ireland Environment Agency, 2009 ).

The Environment Agency has also initiated a consultation, and published draft guidance, on the setting of limits for disposal of liquid and gaseous RAW into the environment under EPR10 (as amended) (Environment Agency, 2011).

The UK government started construction of the Dounreay research station near Thurso in Caithness in 1955 to undertake a major research programme on fast reactor technology (Fig 17.1). The programme was stopped in the mid-1990s as it was considered that fast reactors were not needed in the foreseeable future. The research station was operated by the United Kingdom Atomic Energy Authority (UKAEA) until 2005 when it was taken into ownership by NDA and is currently managed under contract by Dounreay Site Restoration Ltd (DSRL) a wholly-owned subsidiary of Babcock Dounreay Partnership. The area of the licensed site is 57 ha situ­ated within 547 ha of NDA-owned land.

During the 50 years of operations, three large nuclear reactors were built, two fast reactors and a materials research reactor. Each had associated industrial size fuel research and inspection facilities, associated fuel reproc­essing facilities and RAW management facilities (Dounreay, 2012). Both fast reactors also had their own steam generating plant of unique designs that drove a conventional electricity generating plant. The scale of the installations can be seen in Fig. 17.2.

Owing to the wide range of research and the goal of demonstrating the complete fast reactor fuel cycle, which was achieved in the early 1980s, Dounreay has the widest range of radioactive wastes and facilities to manage and the most complex decommissioning challenges of any nuclear site in Scotland. In a UK context, only Sellafield in England has a more hazardous and complex RAW and decommissioning programme.

State regulatory authorities

Provisions of law allow federal agencies to delegate or relinquish certain regulatory responsibilities to the states having radioactive materials or nuclear facilities. NPPs are regulated by federal authorities. Regional arrangements allow closer coordination, such as using radioisotopes for medical uses. These arrangements are not necessarily mandatory; however, where the state can demonstrate adequate competencies, the appropriate federal agency can transfer regulatory authority.

The NSCA and its associated regulations provide the basis for regulatory expectations and decisions. Regulatory documents clarify NSCA require­ments and associated regulations, and are an integral part of the regulatory framework for nuclear activities in Canada. Each regulatory document aims to disseminate objective regulatory information to stakeholders, including licensees, applicants, public interest groups and the public, and promote consistency in the interpretation and implementation of regulatory require­ments. As outlined in the CNSC Regulatory Policy P299, Regulatory Fun­damentals (CNSC, 2005), CNSC sets requirements using appropriate industry, national and international standards. The CNSC regulatory frame­work draws upon Canadian and international standards and best practices, including the nuclear safety standards of the International Atomic Energy Agency (IAEA).

A list of CNSC’s regulatory documents is available online at: nuclearsafety. gc. ca. Two of these documents are specific to the management of RAW. Other more generic regulatory documents that relate to action levels, decommissioning, environmental protection and public information programs may also apply to the management of RAW. The CNSC’s regulatory documents for management of radioactive waste are discussed below.

The CNSC Regulatory Policy P-290, Managing Radioactive Waste (CNSC, 2004) outlines the philosophy and principles used by the CNSC in regulat­ing radioactive waste. The policy considers the extent to which owners of RAW must address:

• waste minimization;

• the radiological, chemical and biological management of RAW;

• the predicted impacts on the health and safety of persons and the environment;

• the measures needed to prevent unreasonable risk to both present and future generations; and

• the trans-border effects on the health and safety of persons and the environment.

The CNSC Regulatory Guide G-320, Assessing the Long Term Safety of Radioactive Waste Management (CNSC, 2006) assists licensees and applicants to assess the long-term storage and disposal of RAW. The guide was developed using provincial, federal and international documents, fol­lowing a consultation with the nuclear industry in Canada.

In addition, the nuclear industry in Canada, in conjunction with the CNSC, has developed two Canadian Standards Association (CSA) stand­ards for the interim management of used nuclear fuel and RAW. These standards incorporate best practices both nationally and internationally. For example, the CSA has developed a standard consisting of best practices for the safe siting, design, construction, commissioning, operation and decom­missioning of facilities and associated equipment for the dry storage of irradiated fuel, known as CSA N292.2-07, Interim Dry Storage of Irradiated Fuel. (CSA, 2007). The standard CSA N292.3-08, Management of Low — and Intermediate-Level Radioactive Waste (CSA, 2008) provides advice on the management of low — and intermediate-level radioactive waste which is based on current best practices, international experience and guidance, and in accordance with the existing CNSC regulatory requirements.