The RB maintains a careful check on the licensee’s activities during opera­tion of the NPP to ensure that the plant is operated within the prescribed safety envelope and that other licensing conditions are complied with. This is done through review of the various operational reports, reports on safety — related incidents and on activities during refuelling outages and other extended outages at the plant. The RB also conducts periodic regulatory inspections and audits of records to physically verify the compliance of license conditions, and to check on the general upkeep of the NPP.

As discussed in Section 20.5.6 above, detailed and comprehensive PSRs of the NPP operation are undertaken by the licensee at specified intervals, typically every 10 years. The RB carefully checks the reports of such reviews to confirm that the NPP is meeting the current safety requirements and is likely to continue to meet them till the next PSR. During all these reviews, the RB should make extensive use of the operating experience from NPPs of similar design as well as other nuclear and conventional industries, as far as is applicable.

The main goal of the surveillance by the authority is to make sure that the operator follows the law and the conditions of the license. Surveillance is carried out by inspectors and assessors within the RB. There may be dif­ferences in the detailed approach undertaken in different countries, or within countries. As an example, data in paragraph A.3 of the Annex describes the activities and efforts undertaken in 2006 in Baden- Wurttemberg, Germany, for which an assessment of the surveillance process is available (ILK, 2006).

A new nuclear programme needs thousands of skilled craftsmen, in addi­tion to engineers, such as welders, boilermakers, iron workers, pipefitters, construction labourers, millwrights, electricians, carpenters, insulators and heavy equipment operators. Hence the importance of vocational schools and apprenticeship programmes.

These specialists require specific training in quality assurance, safety and radiological protection, if they are to work in the nuclear field. The worker is required to achieve nationally accepted standards of competence in order to satisfy the vocational training requirement. The worker’s competence is assessed while observing his or her performance in various standard tasks, by assessing knowledge and understanding (typically by using oral and written questions), and by collecting other evidence about the worker’s competence.

Another approach is through apprenticeship programmes for young stu­dents leaving secondary education who gain, during a certain period of time, real work experience and some complementary nuclear training. The emphasis throughout this training period is on exposing the apprentice to as many of the different facets of nuclear work as possible, while ensuring that the learning process is fully supervised with respect to safety. While set standards are required to be met in order for the apprentice to successfully achieve a formal apprenticeship qualification, this can also be considered as human resources development, using this qualification as the first step in the process of building a successful career in the industry.

Further detailed information regarding SAT methodology can be found in the Institute of Nuclear Power Operations (1993), Principles of Training System Development Manual, ACAD 85-006, as well as several publications from the International Atomic Energy Agency: IAEA (2000), Training solutions / Analysis phase of systematic approach to training (SAT) for nuclear plant personnel, IAEA-TECDOC-1170, and IAEA (2001), A sys­tematic approach to human performance improvement in nuclear power plants, IAEA-TECDOC-1204.

A. ALONSO, Universidad Politecnica de Madrid, Spain

Abstract: This chapter outlines the justification principle and how it could be applied to taking decisions about the development of nuclear power. The justification principle compares the economic, social and environmental benefits derived from a given development against the risks and detriments associated with it. When the benefits outweigh the associated risks and detriments, the intended development is considered to be justified. The benefits from using nuclear power for the generation of electricity arise from its reliability, independence, costs and freedom from carbon emissions, while its risks and detriments are associated with the generation of toxic radioactive products and strategic nuclear materials that need to be kept under control. The justification principle could be used to consider the establishment or continuation of a nuclear development plan, to select an individual design and the corresponding fuel cycle, or to help decide the longer-term operation of an already operating plant.

Key words: benefits of nuclear power, detriments from nuclear power, ethics of justification, justification equation, justification process.

8.1 Introduction

The justification of facilities and activities is the fourth principle of the 10 ‘Fundamental Safety Principles’ introduced by the International Atomic Energy Agency (IAEA) in 2006. The principle states that ‘facilities and activities that give rise to radiation risks must yield an overall benefit’ (IAEA, 2006). This principle was first introduced by the International Commission on Radiological Protection (ICRP) as a basic principle in the protection against ionizing radiation (ICRP, 1990) and has mainly been applied in medical and other uses of radiation.

Many countries have introduced the justification principle into national legislation, although limiting its application to radiation protection in radia­tion uses. The definition of justification given by the IAEA extends the application of the principle to facilities and activities where radiation risks are present; nuclear power plants and related fuel cycle installations and activities lie within this class. Although such installations and activities are clear candidates for application of the principles, many countries have not included such requirements in their general regulatory practices, with the notable exception of the United Kingdom for general applications (UK, 2004) and specifically for nuclear power designs (UK, 2008), as explained in Appendix 1.

Supporters of the application of the principle to the development of nuclear power believe that it serves to balance the benefits and detriments, providing insights into high-level decision processes and aiding the social acceptance of nuclear energy. When solving the justification equation, all possible benefits should be assessed as well as all risks and detriments coming from the construction and operation of nuclear power plants, fuel cycle installations and related activities. Benefits, risks and detriments can be economic, social or environmental. All these elements will be described in this chapter.

The International Nuclear Safety Group (INSAG) advises that the 10 IAEA Fundamental Safety Principles be applied to all the different phases in the life of a nuclear power plant (INSAG, 2008). The application of jus­tification to all phases in the life of a nuclear power plant could be very effective for the early phases, and it should be considered as part of the decision to launch a nuclear power programme and in the selection of acceptable technologies. During plant operation, it could also help when taking decisions such as enlarging the capacity and long-term operation of a nuclear power plant, as well as assisting in extraordinary circumstances, such as when recovering from relevant incidents, and equally could also be used in selecting the decommissioning level and the technology used for it.

INSAG recommends that the justification principle should be applied by new entrants and by those countries interested in expanding their nuclear power programmes (INSAG, 2008). Nevertheless, neither INSAG nor the IAEA have developed detailed technical guidance on how to develop a justification document. The purpose of this chapter is to give such guidance on how to develop the terms included in the justification equation.

The application of the justification principle needs a process and a justi­fication authority. A country’s government is responsible for establishing the regulatory requirements and corresponding guidance, and for selecting the justification authority. Relevant decisions, such as the decision by a state to embark on a nuclear power programme, are generally taken at the highest levels of government. For other decisions, such as for those concern­ing the longer-term operation of existing nuclear power plants, the regula­tory body may determine whether the decision is justified.

The KLT-40S is a pressurized water reactor based on the commercial KLT-40 marine propulsion plant, and is an advanced variant of the reactor plants that power nuclear icebreakers. The construction of a small-size float­ing nuclear cogeneration plant with two KLT-40S reactors is currently under way in Russia. The KLT-40S is a modular design in which the reactor, steam generators and main circulation pumps are connected with short nozzles. It is a four-loop system including forced and natural circulation of the primary coolant, with a once-through coiled steam generator, an exter­nal gas pressurizer system and passive safety systems.

The newness of nuclear generation technology, the dominant place of the designer/builder in development of the system, and the extended process of plant design and construction before the first operating licence is issued tend to leave the impression that the central role is played by this group. However, it must be recognized that the designer/builder leaves the site shortly after first operation and has (at least in Canadian practice) no further responsibility for the plant, following handover from the vendor to an operating organization. Similar handover practices exist in most, if not all, countries. For example, in France where EDF is heavily involved in the building of nuclear plants as well as serving as the sole operating organiza­tion, handover of the plant must be formally executed from the building unit of the company to the operating unit of the same company. The prin­ciple remains the same. The primary role of the designer/builder is to deliver a plant to the operating organization that not only meets regulatory require­ments but also meets the staff and plant protection safety goals. During the operating phase, particularly in the early years, the designer/builder might perform support services to the operating organization. These services must eventually be taken over by either the operating company or a related organization whose only commitment is to support of operating stations. The operating company has a return responsibility to the designer/builder. It must inform designer/builder staff of the design features that are most useful during operation from the point of view of performance and safety, in addition to comprehensive operational feedback on any components or systems that require improvement. Practical considerations will vary in each individual case; in every case the essential linkages that must be sustained

over the whole operating life of the plant are illustrated in Fig 10.1. Fortunately, in recent years it has become ‘best practice’ for the designer/ builder to transfer to the operating organization the same comprehensive CADD (computer-aided drafting and design) model that was used to con­struct the plant. Associated materials lists and other supporting documenta­tion is also transferred to the operating organization, to serve as a complete record of the facility ‘as built’. This model then can be used by the operating organization to maintain a record of all in-service experience and mainte­nance operations. One alternative method for retention of these data is described in another INSAG report (INSAG, 2003).

Programs administered by owners’ groups and national/international orga­nizations such as INPO and WANO are very active in developing codes of good practice for promulgation across the world. These codes represent the best judgment of true experts in the field of nuclear plant operation. These are made available to all members of these organizations; they offer sound support to any newly founded operating organization.

While the radiation protection principles were originally formulated for dealing with protection against ‘certain’ exposures, namely against expo­sures that will occur with some degree of certainty, they may, mutatis mutandi, be used against ‘potential’ exposures as well, namely against situ­ations having the capacity to develop into real exposures in the future. Namely, the principles described heretofore can be used not only for ‘radia­tion protection’ but also for ‘radiation safety’ in general and for nuclear safety in particular. Nuclear safety has been treated in Chapter 10.

Proposals for safety criteria for NPPs founded on the underlying radia­tion protection principles were suggested very early (Gonzalez, 1974, 1982, 1986). The basic proposal was to use available probabilistic assessment tools, such as event and fault trees, for a priori overall safety analyses. A comparison could, therefore, be performed between the probability of occurrence of a hypothetical chain of events leading to an unexpected human exposure, along with its consequences in terms of doses incurred, and a regulatory criterion based on the radiation protection principles. The relevant regulatory authorities would then be able to judge safety levels on the basis of a rational approach sharing the same principles of radiological protection.

A conceptual framework for the protection from potential exposure and how to apply the conceptual framework to selected radiation sources has been recommended internationally (ICRP, 1993, 1997b).

There is at least one practical regulatory application of the radiation protection principles to a nuclear safety criterion (CNEA, 1979, 1980; ARN, 2010), which was discussed at various international meetings (Gonzalez, 1982, 1986). The aim of the regulatory criterion is to require applicants for a NPP licence to identify the failure sequences which, in the case of occur­rence, will deliver a radiation dose to members of the public, and make their probability of occurrence sufficiently low to be coherent and consistent with the radiological protection principles. The probability of occurrence of each failure sequence, as well as the corresponding activity of released radionu­clides, should be assessed by using event and fault tree analyses, which must comply with the following criteria:

1. The failure analysis shall systematically encompass all foreseeable fail­ures and failure sequences, including the common-mode failures, the failure combinations and the situations exceeding the design basis (failure in this context means an aleatory event preventing a component from performing its safety function, as well as any other event which may additionally occur as a necessary consequence of such deficiency; failure sequence, on the other hand, means a sequential series of failures which can, although not necessarily, occur after an initiating event.

2. A failure or a failure sequence may be selected as representative of a group of failures or of failure sequences (in such a case, the failure or failure sequence to be selected from the group shall be that delivering the worst consequences and the analysis shall take into account the sum of the probabilities of the failure or failure sequences in the group).

3. The analysis shall consider that a protection function may have lost operativeness either before the occurrence of the failure or of the failure sequence or as a result of such occurrence.

4. The analyses of failures, of failure sequences or of any part thereof shall be based on experimental data as far as possible (if this cannot be done, the valuation methods must be validated through appropriate tests).

5. The levels of failure rate assigned to the safe-related components, in the evaluation of the failure probability of systems, shall be justified. In the case that justifiable values were not available for some of the compo­nents, the applicant shall use levels of failure rate prescribed by the licensing authority (if a given failure rate is justified on the basis of quality assurance, this must be specified in detail).

6. The failure analyses shall consider the maintenance and testing proce­dures, and the time interval between successive maintenance and testing actions.

7. Failure rates postulated for human actions shall be justified taking into account the complexity of the task, the stress involved and any other factors which might influence that failure rate.

Thus annual probability of occurrence of any failure sequence, when plotted as a function of the resulting effective dose, shall result in compli­ance with a criterion that is coherent and consistent with the principles of radiological protection enunciated above. The implicit basic safety goal is a risk limit derived from the dose limitation system used for radiation protec­tion purposes, which — as seen before — includes four principles: two of them are source-related (e. g. justification and optimization) and the other two are individual-related (e. g. individual limitation and intergenerational pro­tection). These latter principles entail that the risk committed by individual sources should be low enough as to be automatically disregarded. The cur­rently recommended dose limit of 1 mSv per year implies an annual risk limit of around 10-5 for any individual, even for the highest exposed one, as a result of performing all practices involving radiation exposure. However, since the dose limits relate to individuals, appropriate constraints for indi­vidual doses should be selected for each source of exposure. The dose constraint must be sufficiently lower than the relevant dose limit, so as to prevent individual exposure due to several sources from exceeding such limit. Therefore, the de facto annual limit of individual risk would become lower that the limit of around 10-5. On the basis of the above limit and taking into account the uncertainties usually involved in probabilistic safety assessments, an annual risk limit for accidental exposures from nuclear installations should not exceed an order of 10-6. This would be consistent with the principles involved in the currently enforced system of dose limita­tion. Moreover, accidental exposures may arise from a theoretically infinite number of accidental sequences, each one having a given probability of occurrence and delivering a given expected dose to the most exposed indi­vidual. The actual risk incurred by this individual will then result from the integration of the tail distribution of doses (i. e., the complement of the probability function of doses) times the probability of death provided the dose is incurred. The safety constraint should therefore be that the value of this integral be lower than 10-6 per annum.

The assessment of all possible accidental sequences involving radiation exposure is extremely difficult and practically impossible. Therefore, the regulator may be satisfied if around a tenth of the most relevant sequences are identified, assigning them an annual risk limit of 10-7. Since each sequence may result in different doses, a criterion curve may be adopted, which is a relationship between the annual probability of sequence occur­rence and the expected individual dose, each point of the curve representing a constant level of risk. This criterion curve is shown in Fig. 11.7 (Failure of a point to be under the criterion curve does not necessarily mean that the risk constraint is not met, because even in this case, the integral of the tail distribution could be lower than 10-6 annum.)

The logic behind the criterion curve is as follows. For the range of doses from which only stochastic effects of radiation can be incurred, the criterion curve must show a constant, negative, 45° slope in a log annual probability versus log individual dose coordinate axis plane. This would ensure that the annual probability of incurring the dose times the probability of death provided the dose is incurred (the latter being in the order of 10-2 per Sv) will be kept constant. One of the coordinate points in this part of the curve would obviously be {annual probability = ~10-7 annum-1; individual dose = 1 Sv}, because the product 10-5 annum-1 x 1 Sv x 10-2 Sv-1 results in an annual risk of 10-7 annum-1. In the dose range where non-stochastic effects of radiation may occur (i. e., for individual doses higher than around 1 Sv), the slope of the curve should increase in order to take account of the higher risks of death at these levels of dose. For doses higher than approximately 6 Sv, the probability of death approaches unity. From this level to higher doses, the criterion curve should remain constant at an annual probability of 10-7 (because the exposed individual would inevitably die regardless of the level of the dose). Between the coordinate points defined by {annual probability = 10-5 annum-1; individual dose = 1 Sv} and {annual probability = 10-7 annum-1; individual dose = 6 Sv}, the criterion curve should show a shape similar to that of the relationship between the individual dose and

the frequency of death (which, at that range, is approximately S-shaped but, for the sake of simplification, the Authority has decided to approximate these two points by means of a linear-shaped relationship. Finally, the cri­terion curve has been truncated at an annual probability level of 10-2, because the occurrence of incidents having a higher annual probability (regardless of the dose) should reasonably be expected to be unacceptable for any regulator.

It should be emphasized that the criterion curve is individual-related; i. e., it is intended to limit the risk-rate on the individual incurring the highest risk, but does not take into account the overall expected impact from acci­dental situations. The criterion assures a level of safety which is sufficient to ensure that an individual risk constraint, compatible with the philosophy of the dose limitation system, will not be exceeded. It fails, however, to answer positively the old question of the safety engineers, i. e. is such a safety level safe enough as to preclude further safety improvements? An installa­tion complying with the criterion would equally consider whether it is imposing risks (lower than the ‘acceptable’ one) to few individuals, or whether many individuals would incur such risks. If an accident does occur, however, the overall radiological impact will be very different in each case, suggesting that the overall safety level might be lower in the second case than in the first one. Optimization may require further safety improvements in the second case. But, is this really necessary, providing the individual — related criterion is met? And, if so, on what basis can optimization be implemented? These questions are not simple to answer but a logical response would allow for complementing the probabilistic criterion based on individual risk considerations alone.

Radiation protection assessments use the concept of radiation detriment, namely the mathematical expectation of harm, to quantify the impact from a source of radiation exposure. The detriment is an extensive quantity that estimates the combined impact of deleterious effects resulting from expo­sure to a given radiation source. It is defined as the expectation of the harm to be incurred, taking into account the expected frequency and severity of each type of deleterious effect. The detriment incurred by one individual receiving a dose in the range of stochastic effects is proportional to the effective dose incurred, the proportionality factor being the probability that the individual will incur a deleterious effect as a result of the exposure. Therefore, in cases of actual exposures to low levels of dose, the total detri­ment is proportional to the sum of all the individual effective doses incurred,

i. e., to the collective dose commitment (this latter quantity results from the time integration of the collective dose rate, which, in turn, results from the integral of the population spectrum in terms of effective dose rate incurred). It was therefore tempting to use a similar concept for measuring the expected impact from accidental exposures (Beninson and Gonzalez, 1981). For potential accidental exposures, the concept of detriment may keep its theoretical meaning, although it would become a quantity of a second order of stochasticity. In such case, the probability of a given exposure, i. e., the combined probabilities of both an accidental release and an environmental condition (dispersion, deposition), should be introduced in the formulation and integrated over all possibilities. Then, if low doses were expected, the detriment should be proportional to the resulting mathematical expectation of the collective dose commitment. For higher doses, another component of the detriment should be added in order to take into account the non­stochastic effects of radiation.

This idea of using the detriment of a second order of stochasticity, and the related mathematical expectation of collective dose commitment, for quantifying the impact from accidental exposures is really appealing, as the concept would allow for optimizing safety, increasing it to a sufficiently high level that further improvement would not be worthwhile taking into account both the benefits achieved in terms of expected collective dose commitment reduction and the cost of obtaining such reduction. However, unfortunately, it was demonstrated (Beninson and Lindell, 1981) that, at very low prob­abilities, the detriment will lose its usefulness as a basis for decision-making.

In fact, in such cases the standard deviation of the result may be orders of magnitude higher than the actual expectation and the coefficient of variabil­ity would become very large. The detriment is then no longer a central measure of the distribution of harm and, in addition, the uncertainty of the detriment becomes too large to make it meaningful, even if the probability as such could be estimated by safety assessments with an accurate degree of certainty. At very low failure probabilities, the inherent uncertainty of the product of probability and consequences makes the use of this quantity rather doubtful. For these reasons, for potential accidental exposures the principles of justification and optimization are implemented in a less quan­titative manner. The value assigned to the variables follows a utility function of probability and consequence. The utility function usually gives more weight to larger accidents than would be implied by the direct product of probability times consequence.

It must be emphasized that the proposals for using probabilistic safety criteria were never aimed at performing a posteriori ‘confirmatory’ studies of the risk being incurred. Rather, they are aimed to check a priori that the prevention of accidents is coherent and consistent with the radiation protec­tion principles. It should also be underlined that a priori probabilistic analy­sis allows firmly grounded anticipation, when there are frequency data that allow classical statistical treatment, and (with the help of Bayes’s theorem) solidly founded inference when only professional judgement is available.

In sum, an approach to nuclear safety based on the radiation protection principles has a uniqueness: its coherence and consistency vis-a-vis both actual radiation safety situations and potential nuclear safety situations. This exceptionality is at the root of its claim that it is based on a common ethical approach.

13.1.1 Birth of a landmark treaty

On Monday, 16 July 1945, the world’s first nuclear weapon was detonated by the United States. By the mid-1960s, there were five States which had produced and tested nuclear weapons, including China, France, the former Soviet Union (USSR; today called the Russian Federation), the United Kingdom (UK) and the United States (US). Recognizing the negative impact to their respective national interests if other States were to produce and test such devices, two of the nuclear-weapon States, the US and the USSR, sought to erect an institutional mechanism to limit the further spread of nuclear weapons or other nuclear explosive devices. The Treaty on the Non-Proliferation of Nuclear Weapons (NPT) was first adopted on 12 June 1968, and then on 5 March 1970 it was brought into force. There are two distinct but interrelated NPT verification goals: to build confidence between parties; and to deter against treaty violation by risk of detection.

The text of the NPT segregates the signatories into two camps: the ‘haves’ (i. e., nuclear-weapon States, NWS) and the ‘have-nots’ (i. e., non-nuclear — weapon States, NNWS). The Treaty defines a NWS as one which has manu­factured and exploded a nuclear weapon or other nuclear explosive device prior to 1 January 1967. This meant that five States were recognized as declared nuclear-weapon States at the time the Treaty entered into force: China, France, the USSR, the United Kingdom and the United States. Since 1 January 1967, several other States are known to have, or are assumed to have, conducted a nuclear weapons test. These countries are not recognized as nuclear-weapon States according to the NPT’s definition.

The NPT was given an initial 25-year lifespan in Article X of the Treaty, though another provision, Article VIII, entails a review process that occurs every five years with the goal of assuring that the Treaty’s objectives are being realized. During the 1995 NPT Review and Extension Conference, a decision that the NPT shall continue in force indefinitely was included among the package of decisions that were adopted.[12] The Conference also reaffirmed the universality of the NPT, stating ‘Universal adherence to the Treaty on the Non-Proliferation of Nuclear Weapons is an urgent priority. All States not yet Party to the Treaty are called upon to accede to the Treaty at the earliest date, particularly those States that operate unsafeguarded nuclear facilities.’ The most recent NPT Review Conference was held in 2010 and concluded with the adoption of a 22-point action plan (over the next five years) to advance the three main pillars of the Treaty: nuclear disarmament, non-proliferation and the peaceful uses of nuclear energy. The final 2010 document also provides an article-by-article review of the NPT’s operations, taking into account the decisions and resolutions previ­ously adopted by both the 1995 Review and Extension Conference and the 2000 Review Conference.[13]

14.4.1 Overview

As described earlier in Section 14.2.3, there are different ways of looking at spent nuclear fuel, either as a resource to be reprocessed and recycled or as a waste to be disposed of in a geological repository. In the case of reprocessing the valuable materials, plutonium and uranium will be recy­cled as MOX or REPU fuel and the remaining waste will need geological disposal.

The choice between the two options, recycling or disposal, will be based on strategic, political and economic factors. At present about 15-20% of all spent fuel is reprocessed and the plutonium and uranium recycled. The recycling takes place in light water reactors. Although over the years recy­cling has been performed in several countries, it is primarily France that is doing it on an industrial level today. France has both reprocessing and MOX fabrication capacity. Other countries such as Japan, Russia and China are preparing for recycling. Recycling will lead to a better utilization of the natural uranium resource. Recycling in light water reactors will reduce the uranium consumption by about 25%. For more effective use, recycling in fast reactors will be necessary, which will allow multiple recycling of the generated plutonium and which can also utilize the depleted uranium from enrichment, which otherwise would be a waste product, as breeding mate­rial for new plutonium that can be utilized as fuel. Theoretically, recycling in fast reactors could lead to the uranium being utilized at least 50 times more effectively, i. e. one could get 50 times more energy out of the natural uranium. Fast reactors are, however, not yet available on a commercial scale for electricity production. Important development work is going on in several countries, e. g. Russia, France, India, Japan, China and the United States. Except for India, it is, however, not expected that recycling in fast reactors will be of significant importance before 2050. The technical and economic feasibility still needs to be proven before fast reactors can be introduced on a commercial scale. From a waste management point of view there could also be an additional added value in recycling in fast reactors as this could provide the possibility to also burn (transmute into shorter — lived elements that can be disposed of more easily) some of the other transuranic elements, e. g. americium and curium, elements which make an important contribution to the long-term radiotoxicity of the fuel (see Section 14.3.2).

Although the management system for spent nuclear fuel will be different depending on what management strategy is chosen, recycling or disposal, there are also many components in common.

The management system for recycling of the spent nuclear fuel includes the steps shown in Fig. 14.4. Reprocessing is normally performed 3-10 years after the fuel has been removed from the reactor. In principle the fuel can be transported away from the reactor already after about a year, but often the transport is later. At present it is not foreseen to recycle the MOX fuel again in light water reactors, but to store it for later use in fast reactors. It should, however, be realized that the spent MOX fuel has a higher heat generation, radiotoxicity and neutron radiation level than the correspond­ing spent uranium fuel.[86] This will be important should the spent fuel later be considered for direct disposal and not for recycling.

The management system for direct disposal of spent fuel has the compo­nents shown in Fig 14.5. The cooling time before spent fuel can be disposed of in a geological repository will be typically 30-50 years or longer.

Over the years approximately 400,000 tonnes of spent fuel (measured as heavy metal (HM)) have been generated. About 100,000 tonnes of these have been reprocessed and the remaining 300,000 tonnes remain in the reactor pools or are stored in dedicated facilities within the power plant premises or in centrally located storage facilities either for direct disposal or awaiting a later decision to reprocess. More details about different storage facilities at the reactor site or centrally are given in Section 14.4.2.

Except for the reprocessing step, spent fuel management has so far been mainly a national activity. Storage facilities are built at the reactors or cen­tralized in the country. No international storage facilities have been devel­oped. The same is the case for the work on geological disposal. Although there is much international cooperation on research and development for geological disposal, there are no agreements between countries to develop a common geological disposal facility, in spite of the technical/economic advantages it could bear. Discussions have taken place in different fora but no real progress has been seen, mainly due to the political sensitivity of the subject. Reprocessing is the exception. Several countries have jointly financed some reprocessing plants and sent their fuel for reprocessing to these plants. The agreements have, however, included the stipulation that the waste from reprocessing will be returned to the country of origin for further management and disposal.

14.4 Management system for recycling of the spent nuclear fuel.