National policy in the field of RAW management

Major principles of state policy in the field of RAW management are stated in the laws of Ukraine ‘On the Use of Nuclear Energy and Radiation Safety’ and ‘On Radioactive Waste Management’. They are as follows:

• priority to protection of personnel, population and environment against the impact of RAW;

• isolation of RAW from the environment;

• regulatory control of RAW management;

• separation of the regulatory functions and RAW management functions;

• responsibility of RAW producers for its safety;

• decisions on siting of new RAW management facilities taking into account society’s opinions;

• prohibition of transfer to Ukraine of RAW for storage or disposal.

The principles of State policy in the field of spent fuel (SF) management are stated in the law of Ukraine ‘On the Use of Nuclear Energy and Radia­tion Safety’. Energy Strategy (2006) defines the steps in the implementation of the so-called ‘deferred’ decision for SF of Ukrainian NPPs — long-term (up to 50 years and more) storage of SF with subsequent approval of final decision on either processing or disposal.

Radioactive contamination and waste may also arise from accidents. Acci­dents generate radioactive waste of volume and composition which depend on the material involved and the magnitude of the accident. The Interna­tional Nuclear and Radiological Event Scale (INES) was developed in 1990 by international experts convened by the IAEA and the OECD Nuclear

Energy Agency (OECD/NEA) with the aim of communicating the safety significance of events at nuclear installations (IAEA, 2008b). The INES facilitates understanding, using a numerical rating to explain the signifi­cance of nuclear or radiological events in a similar fashion to the Richter scale for earthquakes. INES applies to any event associated with the trans­port, storage and use of radioactive material and radiation sources. Such events can include industrial and medical uses of radiation sources, opera­tions at nuclear facilities, or the transport of radioactive material. Events are classified at seven levels (Fig. 1.6): Levels 1-3 are ‘incidents’ and Levels 4-7 ‘accidents’. These levels consider three areas of impact: people and the environment, radiological barriers and control, and defence in depth. The scale is designed so that the severity of an event is about ten times greater for each increase in level on the scale. Events without safety significance are called ‘deviations’ and are classified Below Scale/Level 0.

The partial core meltdown accident at Three Mile Island (TMI), Penn­sylvania, USA in 1979 was at Level 5 on the INES scale, while those at Chernobyl and Fukushima were Level 7. The proper management of the

1.6 The International Nuclear and Radiological Event Scale (from the IAEA website).

TMI accident meant that there were no person overexposures to radiation and no casualties, so keeping it at Level 5.

Table 1.6 shows the most significant accidents involving radioactive ma­terials. Accidents involving SRS are worryingly common. Over 2,300 cases have been reported of SRS found in scrap metal. A large number of cases have been reported of accidental melting of SRS with scrap metal in, for example, steel and aluminium foundries. The total number of cases of melting SRS with scrap metals exceeds 60 in 18 countries. In Algeciras, Spain in 1998, radioactive gases, aerosols and particles from melting SRS with scrap were released and detected all over Europe. Concentrations up to 2,000 Bq/m3 of 137Cs in the air were detected although the incident had minor consequences. Since 1983, 30 cases of melting of SRS with scrap metal occurred in the US, which required $8-10 million in each case to decontaminate and restore the metallurgical facilities. In 1987 a serious accident occurred in Goiania, Brazil with a 337Cs SRS left within a tele­therapy unit. The SRS was found by two scavengers who took the unit home, removed the source from the unit and ruptured the source capsule. This caused significant contamination of people and the surrounding envi­ronment. Four severely exposed people died and the health of many others was seriously affected. More than 112,000 people were monitored for radia­tion exposure, of which nearly 300 showed 137Cs contamination. The emer-

gency response and clean-up effort of houses, buildings and land lasted six months.

The Chernobyl accident in 1986 was due to lack of care in operation and disregard for standard safety procedures. The resulting steam explosion and fire released about 5% of the radioactive reactor core into the atmosphere. Some 31 people were killed in the first few weeks after the accident, and there have since been other deaths from thyroid cancer due to the accident. An authoritative UN report in 2000 concluded that there is no scientific evidence of significant radiation-related health effects to most people exposed to radiation during or after the accident.

The most recent accident was that of 11 March 2011 at Fukushima in Japan. A major earthquake, followed by a 15 m tsunami caused the deaths of over 20,000 people and led directly to the shutdown of three reactors and eventually to significant escape of radioactive material to the environ­ment. Three of the Fukushima Daiichi reactor cores were severely damaged in the first three days, releasing high levels of radioactive materials into the land, sea and air environments. The Japanese authorities announced an official ‘cold shutdown condition’ in mid-December, as reactor tempera­tures had fallen to below 80°C at the end of October 2011. According to the Japanese government, the total amount of radioactivity released to date is approximately one-tenth that released during the Chernobyl disaster However, the full extent and level of radioactive contamination remain unclear.

Knowledge of physical parameters of solid ‘as generated’ RAW is impor­tant for some processing technologies, like compaction and pyrolysis. The requirement for the content and level of information should come from the facility operator and a methodology to determine the parameters shall be tailored accordingly.

Information on the physical and mechanical parameters of processed waste is substantially more important. Demonstration of key mechanical parameters of a waste form and the entire waste package is usually required by the WAC. This requirement comes from the projected long-term durabil­ity of the waste form (in particular for solidified liquid waste) and also from the design and arrangement of waste packages in the disposal facility, where placement of waste packages in several layers is commonly used. The last requirement is usually solved by use of verified and approved waste con­tainers, providing for sufficient mechanical stability for the entire waste package. Mechanical parameters of the waste form are controlled in the waste producer facility using samples taken during waste processing — the scope of control and methodologies should be developed according to the requirements of the WAC and the expectations/requirements of the disposal site operator.

A. BYCHKOV, Z. DRACE and M. I. OJOVAN, International Atomic Energy Agency (IAEA), Austria

DOI: 10.1533/9780857097446.1.115

Abstract: Technical options for waste streams which arise from nuclear applications, research, power generation, nuclear fuel cycle activities and decommissioning of nuclear facilities as well as NORM-containing waste, are summarized. Since optimal selection of technical options is case specific to the waste management needs, they are not ranked. However, selection criteria for waste processing and disposal technologies are summarized and a systematic approach for selection of optimal solutions is proposed.

Key words: waste management, processing, disposal, classification, categorization, waste routing.

4.1 Introduction

Waste management is a subject that has received considerable attention and is recognized as an important link for public acceptance of nuclear energy and its applications. Technical options and technologies are crucial for safe management of radioactive waste. A wealth of information is cur­rently available about a multitude of waste management technologies and their technically novel and alternative designs, as well as about emerging technologies, which require further development and/or validation. Selec­tion among available options and technologies can be done on a national level, or by waste generators or by waste management organizations. The selection principles may vary by organizational preference, collected or known experience or following an optimization procedure. In any case, because of the costs involved, the potential complexity of technical and environmental considerations, as well as the necessity to ensure adequate performance, the selection mechanism will always require rather clear cri­teria in order to address waste management needs. Some criteria will be fairly general and applicable to almost any waste management system. Others may apply to specific waste categories or to particular waste man­agement steps.

The aim of this chapter is to summarize technical options for waste streams which arise from nuclear applications, research, power generation, nuclear fuel cycle activities and decommissioning of nuclear facilities as well as naturally occurring radioactive materials (NORM)-containing waste and to propose a systematic approach for selection of optimal solutions. IAEA publications [1-3] form the basis for establishing appropriate strategies and infrastructure for the management of radioactive waste. The infrastructure requires selection of an optimized technology/option because of the variety of processes and techniques available for different waste streams at specific waste management steps. The technologies selected for different waste management steps should then be combined in an integrated strategy to optimize the overall waste management system [4] . The selection of waste technologies for each specific waste stream/category should be based on an evaluation process with the following elements:

• identification and nature of specific radioactive waste inventories and associated properties;

• consideration and review of various options for the management of that waste;

• evaluation of the advantages and disadvantages of each option using multi-attribute utility analysis (MUA) [5] or any other suitable method­ology that compares safety, technological status, cost-effectiveness and social and environmental factors;

• selection of the best available technology(ies) not entailing excessive cost and satisfying all regulatory requirements [6];

• approval (via licensing, authorization) of the selected technology(ies).

C. M. JANTZEN, Savannah River National Laboratory, USA, W. E. LEE, Imperial College London, UK and M. I. OJOVAN, University of Sheffield, UK

DOI: 10.1533/9780857097446.1.171

Abstract: The main immobilization technologies that have been demonstrated for radioactive waste disposal are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability. Future waste generation is driven by interest in sources of clean energy. The development of advanced waste forms is a necessary component of the new nuclear power plant (NPP) flowsheets. A brief summary is given of existing and advanced waste forms and processing technologies.

Key words: vitrification, cementation, bituminization, glass, cement, bitumin , waste form, advanced waste forms, conditioning, immobilization, encapsulation.

This section presents the SCC behaviour of carbon steel disposal containers and stainless steel storage canisters. Carbon steel is mostly susceptible to hydrogen-induced cracking due to residual weld (and heat affected zone) stress or seismic-induced impact stress. Hydrogen is generated during general corrosion or gamma radiolysis of groundwater in a reducing envi­ronment (Ahn and Soo, 1995).

In a marine (coastal) environment, salt deposits may occur on the stain­less steel canister surface due to salt deposits in the humid air. The salt deposits on the canister when the canister surface temperature is above ambient. Aqueous conditions of high chloride concentration may form due to this salt deliquescence. With the residual tensile stress at welds (including

7.5 Examples of size and distribution of pits. Pit diameter vs time for 18Cr-12Ni-2Mo-Ti stainless steel in 0.1 n H2SO4 + 0.1 n NaCl. Normal pit size observed is in the range of micrometers to millimeters and pit density is for 304 stainless steel after potentiostatic polarization in 1m NaCl solution. The term ‘d’ is pit diameter and ‘t’ is time (reprinted with permission) (Szklarsksa-Smialowska, 1986). Used with permission from NACE International.

heat affected zone), this high concentration of chlorides may induce SCC (Shirai et al., 2011; EPRI, 2005). SCC can be screened out based on stress mitigation techniques (or other remediation) such as applying compressive stress in the weld. In the absence of this mitigation, the opening surface area by SCC (or hydrogen-induced) cracks may be estimated, if a canister were susceptible to SCC. This quantitative estimate allows assessment of the radionuclide release due to waste form degradation inside the container or the canister.

Y. V. PUZANOV, SUE SIA ‘Radon’, Russia DOI: 10.1533/9780857097446.1.327

Abstract: The axiomatic basics of quantitative safety/risk assessments are discussed. Deterministic and probabilistic analysis methods are then introduced. As an illustrative example, safety assessment for the environment is given in terms of the probability of radionuclide escape from a near-surface disposal facility. Emergency accident levels are correlated with the probabilities of those accidents occurring.

Key words: safety, risk, equipment failure, probability of failure, emergent event tree, failure trees.

9.1 Introduction

Increased attention is being given to issues of safety in the nuclear field as evidenced by the large numbers of publications on this topic. These can be divided into two types according to how the term ‘safety’ is interpreted. The first, the subjective type, concerns the safety of the environment, population or personnel in the proximity of radioactive materials and ionizing radiation sources. The second, the objective type, discusses the safety of nuclear power plants (NPP) or nuclear hazardous objects, although the influence of these objects on the environment or people is also discussed. In the first case, safety is treated in terms of the ability to protect from the effects of ionizing radiation. In the second case, safety is treated as a property of an object, i. e. the property of not rendering an action hazardous, not resulting in contamination or not resulting in the spread of radioactivity. This ambi­guity can lead to misunderstandings in safety assessments. For example, with respect to the ability to protect, it is necessary to take into account the availability/absence of radiation detectors and means of personal protec­tion. If, on the other hand, safety is the property of an object and the need is to avoid hazardous action, taking, for example, the NPP, then other factors must be included associated with the object’s composition, for example the physical barriers designed into the system.

Scientific safety assessment requires that the true character of both sub­jective and objective safety problems is considered quantitatively in math­ematical terms. In this chapter safety assessment is considered using the following system of definitions and axioms:

Definition 1. The safety of an object indicates the state of its immunity from the harmful influence of other objects or factors dangerous to it.

Immunity assumes the ability to resist harmful influence. This ability can be realized through the presence of systems and the elements which prevent penetration of the dangerous object and characteristics of the protected object, i. e. through the presence of a specified protective shield separating the dangerous and protected objects. In other words, it is possible to say that the object is protected if it is supplied with a protective shield (Fig. 9.1).

Definition 2. The protective shield is the means of reducing or eliminating harmful influence on the protected object.

Objects may be divided into the protected or dangerous by analyzing their interaction. The object which has harmful influence is considered as danger­ous. The object, whose safety should be ensured, is protected. For the theory to be developed we must formulate:

Axiom 1. There are protected objects and dangerous objects or factors.

This axiom is needed to clearly establish the safety of the object in question, as it is protected, and dangerous objects may change their roles. For example, if we consider the harmful affects of radioactive wastes (RAW) on the environment, the latter is protected and the wastes are a dangerous object. If we are talking about external influences on disposal wastes in a facility, then all the potential, unauthorized entrants such as terrorists, meteorites, infiltrating moisture, insects and rodents, that is, in fact, the environment, become a dangerous object, and the wastes are what must be protected. This shows that to distinguish the protected objects from the dangerous it is necessary to consider the postulate of their interaction:

Axiom 2. A dangerous object has a harmful influence on the protected.

From Definition 2 and the axioms, the obvious conclusion can be formulated:

9.1 Protective shield.

Safety principle: For the safety of a protected object it should be separated
from the dangerous object or factors by means of a protective shield.

These definitions, axioms and the safety principle are the axiomatic basis of the safety theory as a scientific area. The safety principle enables us to understand that quantitative safety assessments can be made, characterizing the state of the protective shield, its performance and reliability. Therefore, the subject of study of the safety theory is the protective shield. The theory involves the use of mathematical modeling of the system’s functioning as the protective shield and the components that make up the protective shield, as well as numerical methods and natural, physical experiments to determine the characteristics and parameters of these systems.