During reactor operations, the cladding metals, mainly Zircaloy, corrode in water. This introduces hydrogen into the Zircaloy. Hydrogen can degrade the strength of Zircaloy through overall embrittlement caused by a disper­sion of radially oriented hydrides (perpendicular to the hoop stress) (Chung, 2004). The hydrides formed during reactor operations are mostly circum­ferential hydrides (parallel to the hoop stress). Circumferential hydrides may not affect the strength significantly, depending on the magnitude of severity. However, circumferential hydrides are known to become radially reoriented in the presence of appropriate applied stress and temperature (Chung, 2004). Figure 7.12 compares hydrides oriented circumferential or perpendicular to the hoop stress and Fig. 7.13 shows ductility loss with radial hydrides (Yagnik et al., 2004).

Another hydrogen effect is delayed-hydride cracking (DHC). Small cracks that develop on the inner or outer surface of cladding may lead to crack propagation when assisted by hydrogen diffusion to the crack tip, thus forming radially oriented hydrides at the crack tip. The mechanism has not been proven to exist under dry storage conditions. Figure 7.14 shows a schematic for the mechanism of the DHC process. The crack density and size from hydrogen embrittlement of hydride reorientation and DHC can be conservatively assessed like the SCC of stainless steel described in Section 7.4.2.

The RAW isolation process is based on the principle of multi-barrier pro­tection at all stages. RAW conditioning allows the immobilization of radio­nuclides in durable matrix materials (such as glass, cement, ceramics, etc.), which are then placed into special protective casks, such as metal drums or metal and reinforced concrete containers, for transportation. Any voids are filled with special backfill. Conditioned RAW is then placed into hydro — isolating repositories, of which there are three types: near surface, well-type and drill-type. These constructions are supplied with a system of multifunc­tional barriers, which prevents any interaction between RAW and external factors. Interaction between the radionuclides and the environment is pre­vented by both man-made and natural barriers, with each performing its own shielding function. The multi-barrier system means that the safety of the repository is not dependent on one barrier alone, and is assured not only by technical means, but also by technical-organizational measures. The principle of multi-barrier protection guarantees the safe storage of RAW over the whole period during which they pose a radiological hazard.

In accordance with the recommendations of the International Atomic Energy Agency (IAEA), long-term RAW storage facilities must guarantee geo-ecological safety for the entire period of operation. In the national programmes of RAW management, most countries have set the normative period for low and average level activity RAW at 300 and 500 years. This can be explained by the fact that near-surface type repositories are only used for low — and intermediate-level RAW (LLW and ILW) with half-lives of less than 30 years. Over the course of the operation of these repositories, the activity due to natural decay will be lowered, in comparison with the initial level, by three orders for LLW and five orders for ILW.

Inside the Chernobyl Exclusion Zone, RAW management activities are carried out by the State Specialized Companies (SSC) of the Ministry of Emergency: SSC ‘Chernobyl NPP’, SSC ‘Complex’ and SSC ‘Technocentre’. The locations in the Chernobyl Exclusion Zone where radioactive wastes are stored and buried is shown in Fig. 11.2.

RAW management at the ChNPP

SSC ‘Chernobyl NPP’ is carrying out the waste management activities at ChNPP and SO. Collection of ChNPP liquid waste is performed with a pipeline system. Accumulated liquid waste is stored in two tank stores at the ChNPP site. These storage facilities are a system of reservoirs made of corrosion-resistant steel, which are designed to accept 26,000 m3 (liquid RAW storage facility) and 12,000 m3 (liquid and solid RAW storage facility) of waste. This is the low and intermediate level waste: evaporation bottoms, pulp of spent ion exchange resin and pearlite pulp. Spent radioactive oil is also held in temporary storage tanks.

There are two storage facilities for solid waste at ChNPP site. The first storage facility is a surface concrete structure, which is divided into three groups of compartments for storage of LLW, ILW and HLW. The capacity of compartments is 1,087 m3,1,005 m3 and 1,884 m3 respectively. The second storage facility comprises 26 compartments for LLW, ILW and HLW. Its total volume is 10,000 m3. Waste storage facilities are equipped with special protection systems.

Low and intermediate solid waste, generated as a result of work on ter­mination of operation of power units and activities to transform the ‘Shelter’ object into an ecologically safe system, are collected and moved to the RWDP ‘Buriakivka’. At the same time, high level solid waste is collected in the primary steel 200 L containers. Primary containers are inserted into shielding containers weighing about 4,000 kg, made of steel-reinforced con­crete. Shielding containers are placed in a special temporary storage at the ChNPP site.

SSC ChNPP continues construction of facilities for RAW management with international financial support. These are the Liquid Radioactive Waste Treatment Plant and Industrial Complex for Solid Radioactive Waste Management (ICSRWM). ICSRWM includes:

• Lot 0 — Interim storage of low and intermediate level long-lived and HLW, which is constructed inside the ChNPP building.

• Lot 1 — Facility for removal of solid waste from their stores.

• Lot 2 — Solid waste processing plant.

• Lot 3 — Near-surface storage facility for solid waste at the site of the ‘Vector’ complex.

Exempt wastes are the wastes that meet the criteria for clearance, exemp­tion or exclusion from regulatory control for radiation protection purposes as described in Refs [4] and [5]. The concentration of radionuclides in exempt wastes is negligibly small and no provisions are required for radia­tion protection of professional staff and public, irrespective of the disposal route (RAW disposal facilities or common conventional landfills). No special requirements are established for management and disposal of EW, and for RAW managers and for technologists it is always questionable whether EW should be considered in RAW management planning, or whether such wastes could be omitted and managed as non-radioactive waste. The exemption procedure, consensual criteria for exempt waste and exemption levels for total activity and activity concentration of individual radionuclides are established in Refs [4] and [5]. They are based on dose rates for the public, recommended by the ICRP and generally accepted worldwide.

The first few requirements are directed at governments and require appro­priate national legal and regulatory frameworks to be established within which RAW management activities can be planned and safely carried out. This includes the clear and unequivocal allocation of responsibilities, the securing of financial and other resources, and the provision of independent regulatory functions. Consideration also has to be given to providing pro­tection beyond national borders as appropriate and necessary for neigh­bouring countries that may be affected. Governments must also ensure that a national policy and a strategy for RAW management are established that are appropriate for the nature and amounts of RAW in the country. They must indicate the regulatory control required for particular RAW manage­ment facilities and activities, and be compatible with any regional or inter­national conventions and codes that have been ratified by the country. The national policy on radioactive waste management and strategy to imple­ment it must then form the basis for decision making with respect to the management of RAW within the country.

The regulatory body needs to establish regulations for the development of RAW management facilities and activities and to set out procedures for meeting requirements for the various stages of the licensing process. It has to review and assess the safety case for RAW management facilities and activities prepared by the operator both prior to authorisation and periodi­cally during operation. Provisions must also be in place for issuing, amend­ing, suspending or revoking licences, subject to any necessary conditions and the regulatory body has to carry out activities to verify that the opera­tor meets these conditions.

Operator organisations have the prime responsibility for safety and are required to carry out safety assessments and develop a safety case demon­strating safety. They must also ensure that the necessary activities for siting, design, construction, commissioning, operation, shutdown and decommis­sioning are carried out in compliance with legal and regulatory require­ments. Interdependences among all steps in the pre-disposal management of RAW, as well as the impact of the anticipated disposal option have to be appropriately taken into account and the regulatory authorities must ensure this in the event of different operator organisations having responsibility for different aspects of waste management such as treatment, transport, storage and disposal. An integrated approach must also be taken to both safety and security in the pre-disposal management of RAW. The quality of all work influencing safety must be of a high standard and in this regard appropriate management systems must be applied for all steps and ele­ments of the work undertaken.

All RAW has to be identified and controlled and the amount of RAW arising needs to be kept to the minimum practicable. At various steps in the pre-disposal management of RAW, the RAW has to be characterised and classified in accordance with requirements established or approved by the regulatory body.

All radioactive material for which no further use is foreseen, and with characteristics that make it unsuitable for authorised discharge, author­ised use or clearance from regulatory control, has to be processed as radioactive waste. The processing of radioactive waste needs to be based on appropriate consideration of the characteristics of the waste and of the demands imposed by the different steps in its management (pre-treatment, treatment, conditioning, transport, storage and disposal). Waste packages need to be designed and produced so that the radioactive material is appropriately contained both during normal operation and in accident conditions that could occur in the handling, storage, transport and disposal of waste.

Waste is to be stored in such a manner that it can be inspected, monitored, retrieved and preserved in a condition suitable for its subsequent manage­ment with due account taken of the expected period of storage. To the extent possible, passive safety features must be applied in the design and operation of storage facilities. For long-term storage in particular, measures need to be taken to prevent degradation of the waste containment. Waste packages and unpackaged waste that are accepted for processing, storage and/or disposal must conform to criteria that are consistent with the safety case.

The safety case for RAW management facilities and activities is of high importance and operators have to prepare a safety case and a supporting safety assessment, which must also be reviewed and updated from time to time as circumstances evolve. The safety case must include a description of how all the safety aspects of the site, the design, operation, shutdown and decommissioning of the facility, and the managerial controls satisfy the regulatory requirements. It must also demonstrate the level of protection provided and provide assurance to the regulatory body that safety require­ments will be met. The safety case and its supporting safety assessment have to be documented at a level of detail and quality sufficient to demonstrate safety, to support the decision at each stage and to allow for independent review and approval. Documentation has to be clearly written and include arguments justifying the approaches taken in the safety case on the basis of information that is traceable.

Waste management facilities must be located and designed so as to ensure safety for the expected operating lifetime under both normal and possible accident conditions, and for their decommissioning. They need to be con­structed in accordance with the design as described in the safety case and approved by the regulatory body, and commissioning needs to be carried out to verify that the equipment, structures, systems and components, and the facility as a whole, perform as planned. Facilities have to be operated in accordance with national regulations and with the conditions imposed by the regulatory body. Operations need to be based on documented pro­cedures and due consideration given to the maintenance of the facility to ensure its safe performance. Emergency preparedness and response plans, if required to be developed by the operator, have to be subject to the approval of the regulatory body. Operators have to develop, in the design stage, an initial plan for the shutdown and decommissioning of the pre­disposal RAW management facility and periodically update it throughout the operational period. The decommissioning of the facility has to be carried out on the basis of the final decommissioning plan, as approved by the regu­latory body. In addition, assurance must be provided that sufficient funds will be available to carry out shutdown and decommissioning.

Some facilities are subject to agreements on nuclear material accounting (nuclear safeguards), and in the design and operation of such facilities the system of accounting for, and control of, nuclear material needs to be imple­mented in such a way as not to compromise the safety of the facility.

The requirements set out above are aimed at new facilities, but some existing facilities were not developed to such standards and in such cases their safety needs to be reviewed to verify compliance with requirements. Safety related upgrades need to be made by the operator in line with national policies and as required by the regulatory body.

The byproducts of the operation of fission reactors can be managed in two ways: without treatment that supports recycle or with the recycle option enabled. In the former approach (once-through or open cycle), the fissile resource is the 235U taken from the ground (enriched to 3-5% to allow light water moderation of neutrons) and the several percent of 238U that is trans­muted to 239Pu and fissioned during normal reactor operations. The used fuel once removed from the reactor is considered as waste and is managed accordingly. The latter option (closed loop) can be pursued with varying degrees of processing, though at present only plutonium isotopes are recy­cled in mixed oxide (MOX) fuel (while the residual 235 236 238u is stored for future use).

Though plutonium production reactors and reprocessing facilities oper­ated within the nuclear weapons complex in the US between 1944 and the early 1990s, the reprocessing of spent fuel from commercial reactors was practiced for only a very brief period during the 1970s. A de facto morato­rium was placed on reprocessing of commercial spent nuclear fuel in the US in 1977; this ban was lifted in 1981, but no attempts were made to revisit this option until recently. A 2005 energy bill has again allowed consideration of more complete used fuel management and has spurred a modest revival of research into reprocessing (and transmutation) options. However, current US nuclear fuel management policy remains the once-through option with direct disposal in a deep geological repository.

Most nuclear power producing nations practice the once-through option; France, Japan, the UK and Russia operate at least partially closed fuel cycles in which fuel grade Pu is recovered for recycle. At present, no country operates a more complete recycling program, though research exploring options is in its third decade. It can easily be understood that the waste management issues associated with these options are markedly different (although ultimately it is generally accepted that every option will require a geological repository for the residual radioactive materials).

Use of the FBSR process to produce a highly leach resistant mineralized waste form from Hanford low activity waste (LAW) has been investigated since 2001 (see page 217). Initial studies focused on producing and testing the granular mineral product created by processing high sodium waste feeds with clays at ~720°C to produce nepheline (NaAlSiO4) and nepheline-based minerals such as the sodalites to host I, F, Cl, and nosean to host sulfate and sulfide. Numerous studies (74-80) have shown that it is possible to produce a mineral waste form that effectively immobilizes both radionuclides and hazardous constituents.

To be accepted for near-surface disposal, the waste form is required to meet an acceptance criterion for compressive strength of 500 psi. This requirement is derived from a Nuclear Regulatory Commission Branch Technical Position on low level waste (LLW) forms in the US, which some­what arbitrarily specifies 500 psi to preclude subsidence in the waste dis­posal system. It is also noted that a monolithic waste form reduces the impact to human health for the intruder scenario in the waste site perform­ance assessments. While a monolith is desirable, there are other means by which this requirement can be met, e. g. waste stabilization in high integrity containers (HICs).

In 2005-2006 the Savannah River National Laboratory (SRNL) per­formed a monolith feasibility study for granular FBSR product [157]. Mon­oliths were made out of ordinary Portland cement (OPC) at 80-87 wt% FBSR loading, out of ceramicrete (a blend of MgO and monopotassium

Na2O

6.4 Formulation region for geopolymers compared to hydroceramics in the Na2O-SiO2-Al2O3 (mol%) ternary. Note that the fourth dimension is water content and not shown on the ternary mol% diagram. The geopolymer region labeled as G1 is the target range. Optimum formulations are designated as A and B and a 1" x 2" cylindrical monolith made with composition A is shown in the photograph.

phosphate (KH2PO4)) at an FBSR loading of 35.7 wt%, and out of hydroce­ramics (aluminosilicate zeolite phases formed from metakaolin plus NaOH) at FBSR loadings of 50-80 wt%. The hydroceramics had the best durability as they had a similar chemical makeup to the FBSR product (see Fig. 6.4) but the hydroceramics required hydrothermal processing. Therefore, geopolymers were used to bind the granular mineral waste form due to the similarity of the chemical makeup (see Fig. 6.4) to the FBSR product and the fact that the geopolymers did not require hydrothermal processing. Up to 70 wt% granular product was stabilized in the geopolymer. The granular mineral stabilized geopolymer was shown to be more durable than the granular product alone [158].

Release criteria determine the radiological end-state of a remediated site and consequently the extent of remediation works and the amounts of waste generated. The IAEA have published a safety standard on release of

sites from regulatory control on termination of practices (IAEA, 2006b). Further guidance and examples are given in a Technical Document (IAEA, 2012). The safety standard represents good practice from within the IAEA member states and can be used as a guide by states when establishing their arrangements and regulations. Regulating the release of sites is a national responsibility.

The IAEA standard uses the term ‘practice’ to refer to any human activ­ity that introduces additional sources of exposure or exposure pathways to people. The standard applies to cases where there is a proposal to release sites from the requirements for radiation protection of the appropriate regulatory body because practices have ceased.

The IAEA standard establishes an approach whereby target dose criteria are compared to a prospective effective dose assessment for a critical group of the public, above the pre-practice background levels, of that dose received after the site has been released for defined new uses. The dose assessed is the summed effective dose arising from the land, buildings and other sources that remain at the point of site release or licence termination.

The radiological protection principles of justification, dose limitation and optimization apply to decommissioning and are carried through to site release. The standard recommends a dose constraint for the released site of less than 300 microsieverts per year and a limit below which further dose reduction measures are unlikely to be warranted of approximately 10 microsieverts per year. The zone between 10 and 300 microsieverts per year is considered to be a zone of optimization for both restricted and unre­stricted land release.

The IAEA approach provides both for cases where the release is without restrictions and for cases where the future uses of the land remain under some form of use restrictions. In the restricted case, it is possible to carry out prospective effective dose assessments with assumptions that certain sources remain under control and hence a greater degree of residual con­tamination can remain in situ. The standard recommends that should such controls fail, the effective dose should not exceed 1 millisievert per year.

The IAEA standard describes a generic approach to site release and licence termination which is expanded upon and developed within this chapter as a whole. The dose criteria for release are developed through evaluation of potential radiological consequences through all relevant exposure pathways into radionuclide release criteria (Becquerels per gram). These can be determined generically and set by the regulatory body or be developed on a case-by-case basis. Generic criteria may be more conserva­tive because of the need to make generic assumptions in the dose assess­ment. The released site should be assessed for a variety of exposure scenarios including those in which material from the site is reused or circulated outside of regulatory control. The assessment should take into account uncertainties such as those arising from sampling and analysis.

The standard describes the roles of the national government, the regula­tory body and the operator in site release and licence termination. The national government should establish a legal framework under which ter­mination of practices can occur. The regulatory body should establish detailed criteria and associated guidance, review submissions for site release, perform inspections, take actions if required and issue the licence termina­tion once due process has been completed. The operator is responsible for safe completion of decommissioning, remediation, clean-up and licence termination processes under a specific management system. The management system should cover a process for licence termination, respon­sibilities, competency, calibration and maintenance of survey equipment, quality assurance, record keeping, independent assessment/auditing and non-conformance.

As an example of national approaches, the UK regulatory approach is given in the following. In the UK, licence termination is referred to as ‘de­licensing’. Major nuclear facilities are licensed under the Nuclear Installa­tions Act 1965. Prior to 2005 several examples of de-licensing occurred on parts of UK nuclear sites and for some small research facilities. In 2005, with progress towards large-scale decommissioning in the UK, the Health and Safety Executive (HSE), as the principal regulator, issued formal cri­teria for de-licensing (HSE, 2005). The main features are: [24]

the risk criterion. Where these generic values are not used, a specific case-by-case risk assessment may be submitted by the operator.

• Where practicable, sources of ionizing radiation (e. g., a radiographic source) should not be within the de-licensing area at the time of de­licensing (but may be returned later where this does not require a nuclear licence). Materials which could be defined as radioactive waste (RAW) under UK legislation should not be present on the site.

The UK de-licensing regime follows the main features of the generic inter­national arrangements suggested by the IAEA. The UK uses the lower of the range of optimization suggested by the IAEA for unrestricted use and has no arrangements for licence termination under restricted uses. Several significant parts of nuclear sites have been de-licensed using these criteria since 2005.

It is often the case that regulatory authorities take a conservative approach to release of areas after termination of practices because the decision rep­resents a point at which formal control is relinquished. In many cases the most conservative dose criteria are used and the assessments are based upon very extensive site investigation. In the UK, for example, a 1 in a million per year risk target is used to achieve unrestricted release. For some sites with complex and extensive histories, it may not be practicable to achieve such rigorous criteria without tremendously costly clean-up and very large waste production. For these sites the concept of restricted reuse under ongoing regulatory controls less onerous than full licensing as required for the original practice should be an approach more widely employed. This may require in-situ waste ‘disposal’ authorizations or other forms of institutional control. The restricted release approach may be par­ticularly applicable where the next use of the land is for industrial or new nuclear uses for which the potential for public exposure is limited.

If a pragmatic approach to release after termination of practices is not taken by all parties, there is potential for a ‘greenfield’ approach in which all physical assets are removed and the site returned to an essentially virgin state in order to enable release. In many cases it will be more appropriate to recognize the value of existing buildings, assets and infrastructure and attempt to retain these through the termination process for economic reuse.

The various relevant aspects of activity are separately regulated at the legislative level by federal laws such as ‘On the Use of Atomic Energy’, ‘On the Radiation Safety of the Population’, and ‘On the Sanitary-epidemiolog­ical Prosperity of the Population’, among others. Russia’s ratification of the united convention on the safe management of RAW and spent fuel shows that there is a general trend towards the creation and further development of the national normative lawful regulation of activity with regard to RAW management.

RAW treatment is currently regulated by the following standards and rules:

• ‘ Safety regulations regarding the rotation with radioactive wastes of atomic stations’ (NP-002-04),

• ‘ Collection, processing, storage and conditioning of liquid radioactive wastes. Requirements of safety’ (NP-019-2000),

• ‘ Collection, processing, storage and conditioning of solid radioactive wastes. Safety requirements’ (NP-020-2000),

• ‘Handling gaseous radioactive wastes. Safety requirements’ (NP-021-2000),

• ‘Health regulations for treatment with radioactive wastes’ (SPORO-2002).

The regulations in these documents are applicable to all nuclear facilities, radiation sources and RAW processing units, whether planned, in prepara­tion or operational. The set of documents listed above corresponds to the IAEA’s recommendations concerning the regulation of RAW management.

The transportation of RAW and RAM is governed by health and safety regulations regarding:

• the transportation of radioactive materials (NP-053-04),

• RAW treatment (SPORO-2002),

• The radiation safety of staff and population during transportation of RAM (substances) (SP 2.6.1.128 1-03).

These documents present the established principles of RAW and RAM transportation, and the requirements put in place to ensure the safe trans­port of RAW and RAM.

RAW storage is regulated by:

• ‘Rules on safety provision during the temporary storage of radioactive wastes, which are formed during the output, processing and use of min­erals’ (NP-052-04),

• ‘Collection, processing, storage and conditioning of LRAW. Safety Requirements’ (NP-019-2000),

• ‘Collection, processing, storage and conditioning of SRAW. Safety Requirements’ (NP-020-2000),

• Health regulations regarding RAW treatment (SPORO-2002).

Safety must be ensured during RAW storage in order to prevent staff, the general population and the environment from being exposed to radia­tion over the established limits both under normal operating conditions and in emergencies.

At the RAW disposal stage, safety measures aim to ensure reliable isola­tion of RAW, which in turn ensures radiation safety of the population and the environment for the whole period during which the RAW poses a potential hazard. The principal regulations regarding RAW disposal are:

• ‘Radioactive waste disposal. Principles, criteria and basic safety require­ments’ (NP-055-04),

• ‘Near-surface RAW disposal. Safety requirements’ (NP-069-06),

• ‘ Recommendations regarding the establishment of the criteria of the acceptability of conditioned RAW for their storage and disposal’ ([RB]-023-02),

• ‘ The safety evaluation of the near-surface repositories of radioactive wastes’ (RB-011-2000);

• Health regulations regarding RAW treatment (SPORO-2002).

These documents establish the principles, criteria and basic safety require­ments relating to near-surface RAW disposal, disposal into deep geological formations, and also for LRAW disposal. They establish a classification of near-surface disposals for RAW and make recommendations regarding safety evaluation methods for near-surface repositories. The regulations treat the methods used for RAW conditioning as a basic step in the prepa­ration of RAW for storage and disposal.

The annual generation of low and intermediate liquid level waste from Polish research reactor operation ranges from 30 to 160 m3. The liquid waste is subjected to an evaporation process or is purified by sorption. The evapo­rator bottom concentrates are solidified by bituminisation or cementation and then disposed of at the near-surface type central repository at Rozan. Annual production of solid waste is in the range 5-20 m3 . Solid waste is compacted into carbon steel zinc-plated drums and then transported to the Roz an repository.

About 90% of the liquid waste originated from the Maria reactor opera­tion, while the rest comes from radioisotope production or after decontami­nation. Annual production of the solid waste from industry, hospitals or research activities is in the region of 15-40 m3, spent sealed sources of about 1,000 pieces and smoke detectors of about 20,000 pieces. There is some waste after uranium mining activities which took place in Lower Silesia in the south-west of the country which ended in 1968. There are some 100 dumps of waste rock and ore totalling approximately 1.4 x 106m3.