IAEA-TECDOC-1639 Rev.1
Assessment of Nuclear Energy Systems Based on a Closed Nuclear Fuel Cycle with Fast Reactors
A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)
(Companion CD-ROM)
EDITORIAL NOTE
This CD-ROM has been prepared from the original material as submitted by contributors. Neither the IAEA nor its Member States assume any responsibility for the information contained on this CD-ROM.
The use of particular designations of countries or territories does not imply any judgement by the publisher, The IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. FOREWORD A Joint Study was started in 2005 and completed in 2007 within the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). Canada, China, France, India, Japan, the Republic of Korea, the Russian Federation, and Ukraine participated in this study. The objectives were to assess a nuclear energy system based on a closed fuel cycle (CNFC) with fast reactors (FR) regarding its sustainability, determine milestones for the nuclear energy system deployment, and establish frameworks for, and areas of, collaborative R&D work. The assessment was carried out in accordance with requirements of INPRO methodology and guiding documents of the Joint Study developed and approved by the participating parties (Canada and Ukraine participated in the discussions during the Joint Study but did not contribute to the assessments themselves). The Joint Study was implemented in steps. In its first step, nominated experts in course of extensive discussions analyzed the country/region/world context data, discussed national and global scenarios of introduction of the INS CNFC-FR, identified technologies suitable for the INS, and arrived at a broad definition of a common INS CNFC-FR. In the second step, the participants of the study examined characteristics of INS CNFC-FR for compliance with criteria of sustainability developed in the INPRO methodology in the domain of economics, safety, environment, waste management, proliferation resistance, physical protection and infrastructure. The results of the study were submitted to and endorsed by the INPRO Steering Committee meetings held in Vienna 2005−2007. The authors of the report highly appreciate the valuable comments provided by delegates of INPRO Steering Committee meetings as well as the advice and assistance of the other experts. Due to the length of the Joint Study report a summary of the results was produced, which was published as a hard copy. The full text of the Joint Study report is available on this CD-ROM. The IAEA officer responsible for this publication was R. Beatty of the Division of Nuclear Power, Department of Nuclear Energy.
CONTENTS
SUMMARY ...... 1
CHAPTER 1. OVERVIEW OF THE ASSESSMENT RESULTS ...... 6 1.1. Scope of designs and technologies under consideration ...... 6 1.2. Multidimensional assessment of the INS ...... 7 1.2.1. Safety ...... 7 1.2.2. Environment and resource sustainability ...... 8 1.2.3. Waste management ...... 9 1.2.4. Proliferation resistance ...... 10 1.2.5. Infrastructure ...... 11 1.2.6. Economics ...... 12 1.3. Overall judgment ...... 13
CHAPTER 2. OUTLINE OF THE JOINT STUDY ...... 15 2.1. Position of the Joint Study in INPRO ...... 15 2.2. Main objectives ...... 16 2.3. Organizational features and management of the project ...... 16 2.4. Development of the INS CNFC-FR ...... 19 2.5. Global energy scenarios ...... 19 2.6. National nuclear scenarios ...... 20 2.7. Objectives accomplished ...... 22 2.7.1. Consolidation of outlooks ...... 22 2.7.2. Identification of a reference CNFC−FR for the INPRO assessment ...... 22 2.7.3. Assessment of the INS CNFC−FR ...... 22 2.7.4. Identification of collaborative R&D work and priorities for possible cooperation ...... 24 2.8. Objectives partly accomplished and not accomplished ...... 26 2.9. Feedback to the methodology improvement ...... 27 2.9.1. INPRO methodology as a step in developing a comprehensive instrument for INS assessment...... 27 2.9.2. Identified ways for the methodology improvement ...... 27 2.10. Added value ...... 29
CHAPTER 3. NATIONAL NUCLEAR SCENARIOS ...... 30 3.1. China ...... 30 3.1.1. Brief history and current status of the fast reactor technology development ...... 30 3.1.2. Incentives and boundary conditions for the future deployment of INS CNFC-FR ...... 31 3.1.3. National planning of capacities and structure of energy sector ...... 34 3.1.4. The optimistic and moderate scenarios for development of the CNFC with fast reactors...... 35 3.1.5. National approach to identifying fast reactors designs and relevant fuel cycles ...... 36 3.2. France ...... 40 3.2.1. National strategies and scenarios for development of a closed fuel cycle with fast reactors (CNFC−FR) ...... 40 3.2.3. The transition scenario from present PWRs to fast reactors in France ...... 48
3.2.4. Main trends of R&D efforts for fast reactor development in France ...... 50 3.3. India ...... 53 3.3.1. Planning of capacities and structure of nuclear power sector ...... 53 3.3.2. Brief history and current status of fast reactor technology development .... 54 3.3.1. Incentives and boundary conditions for future deployment of INS ...... 56 3.3.4. Scenarios for deployment of a closed nuclear fuel cycle with FBR ...... 57 3.3.5. Fast reactor design approach with a closed fuel cycle ...... 58 3.4. Japan ...... 62 3.4.1. Brief history and current status of fast reactor technology development .... 62 3.4.2. Incentives and boundary conditions for the future deployment of INS CNFC−FR ...... 63 3.4.3. National planning of capacities and structure of energy sector with determination of share of nuclear power and fast reactor ...... 66 3.4.4. Optimistic and moderate scenario for development of a closed nuclear cycle with fast reactors ...... 67 3.4.5. National approach identifying fast reactors designs and relevant fuel cycles ...... 69 3.5. Republic of Korea ...... 72 3.5.1. National planning of capacities and structure of energy sector with determination of the share of nuclear power and fast reactors ...... 72 3.5.2. Brief history and current status of the fast reactor technology development ...... 74 3.5.3. Scenarios for development of a closed nuclear cycle with fast reactors ...... 75 3.5.4. National approach to identifying fast reactors designs and relevant fuel cycles ...... 81 3.6. The Russian Federation ...... 81 3.6.1. Brief history and current status of fast reactor technology development .... 81 3.6.2. Incentives and boundary conditions for deployment of INS based on CNFC−FR ...... 95 3.6.3. National planning of capacities and structure of energy sector with determination of the share of nuclear power and fast reactors ...... 97 3.6.4. The optimistic and moderate scenarios for development of a closed nuclear fuel cycle with fast reactors ...... 98
CHAPTER 4. COMMON AND SPECIFIC FEATURES IN NATIONAL APPROACHES ...... 103 4.1. Evolution of targets for CNFC−FR development ...... 103 4.2. Objectives, milestones and time frames of national programmes ...... 105 4.3. Scope of designs and technologies under consideration ...... 107 4.4. Preconditions favouring collaborative efforts ...... 110
CHAPTER 5. GLOBAL CONTEXT OF THE ASSESSMENT ...... 116 5.1. Common vision and partnership ...... 116 5.2. Discussion of scenarios ...... 117 5.3. Harmonization of national and global scenarios ...... 119 5.3.1. Assurance of uranium fuel supply ...... 119 5.3.2. Assurance of plutonium fuel supply ...... 121 5.4. Generalized models for global energy mix simulation ...... 123 5.5. Reference model of INS CNFC−FR ...... 124 5.5.1. Sodium cooled pool type reactor ...... 124
5.5.2. Fuel cycle plants ...... 127 5.5.3. Experience and knowledge ...... 130
CHAPTER 6. ECONOMICS ...... 133 6.1. Introduction ...... 133 6.2. Consideration of the assessors ...... 133 6.3. Input for assessment: macro-economic and boundary conditions ...... 134 6.3.1. Power systems and energy resources ...... 134 6.3.2. Electricity markets ...... 135 6.3.3. Nuclear power, reactor types, fuel cycle schemes ...... 139 6.3.4. National deployment scenarios ...... 140 6.4. INS description ...... 140 6.4.1. Reference INS CNFC−FR ...... 140 6.4.2. Peculiarities of national INSs...... 140 6.5. Results of INS CNFC−FR economics assessment ...... 141 6.5.1. Cost competitiveness ...... 141 6.5.2. Ability to finance ...... 145 6.5.3. Risk of investment ...... 146 6.5.4. Conclusion on economic assessment ...... 149 6.6. Feedback on the methodology improvement ...... 150
CHAPTER 7. ENVIRONMENT AND RESOURCE SUSTAINABILITY ...... 151 7.1. Introduction ...... 151 7.2. Approach used ...... 152 7.2.1. Nuclear scenario neutronics calculation ...... 152 7.2.2. Life cycle analysis...... 153 7.2.3. Impact studies ...... 153 7.2.4. Discussion of the indicators ...... 154 7.2.5. Scenarios description ...... 155 7.3. Results ...... 156 7.3.1. Main conclusions on acceptability of expected adverse environmental effects induced by the assessed INS (INPRO environmental basic principle BP1) ...... 158 7.3.2. Main conclusions on resource sustainability and fitness for purpose (INPRO environmental basic principle BP2) ...... 160 7.4. Current limitations of the approach used ...... 160 7.5. Feedback on the methodology ...... 161
CHAPTER 8. SAFETY OF NUCLEAR INSTALLATIONS ...... 163 8.1. Introduction ...... 163 8.2. Safety Features and Challenges ...... 164 8.2.1. Safety features ...... 164 8.2.2. Challenges ...... 164 8.3. Reactor Safety Philosophy ...... 165 8.4. Robust Design Features ...... 165 8.4.1. Innovations towards enhanced safety in the core ...... 165 8.4.2. Sodium fire and sodium water reactions ...... 166 8.4.3. Reliable and diverse shutdown systems ...... 166 8.4.4. Decay heat removal systems ...... 166 8.4.5. Core catcher ...... 166
8.5. Probabilistic safety analysis ...... 167 8.6. Safety of Fuel Cycle Facilities ...... 167 8.6.1. Fuel fabrication ...... 167 8.6.2. Fuel reprocessing facilities ...... 168 8.6.3. Alternate processes ...... 169 8.6.4. Human factors ...... 170 8.7. Assessment of Safety in CNFC−FR by INPRO Methodology ...... 171 8.7.1. Assessment of reactor safety ...... 171 8.7.2. Assessment by participants ...... 171 8.7.3. Assessment of safety of fuel cycle facilities ...... 172 8.7.4. Innovations for future sodium cooled fast reactors ...... 173 8.7.5. R&D in reactor safety proposed by various participants ...... 173 8.7.6. R&D studies on safety of fuel cycle facilities ...... 174
CHAPTER 9. WASTE MANAGEMENT ...... 176 9.1. Introduction ...... 176 9.2. Waste minimization ...... 176 9.3. Protection of the human health and the environment ...... 180 9.4. Waste management to avoid burden on future generations ...... 181 9.5. Waste optimization ...... 182 9.6. Research and development ...... 183
CHAPTER 10. PROLIFERATION RESISTANCE ...... 185 10.1. Introduction ...... 185 10.2. Proliferation resistant fuel cycle ...... 185 10.2.1. Fast reactors ...... 186 10.2.2. Closed fuel cycles ...... 187 10.3. Assessment of proliferation resistance of INS ...... 188 10.4. Feedbacks for Improvement of Methodology ...... 190
CHAPTER 11. INFRASTRUCTURE ...... 192 11.1 Introduction ...... 192 11.2. Description of the case assessed ...... 192 11.3. Discussion of assessment results ...... 193 11.3.1 INPRO user requirement UR1 legal and institutional infrastructure ...... 194 11.3.2. INPRO user requirement UR2 economic and industrial infrastructure ...... 194 11.3.3 INPRO user requirement UR3 socio-political infrastructure (public acceptance) ...... 196 11.3.4. INPRO user requirement UR4 human resources ...... 197 11.4. Feedback to INPRO methodology ...... 199 11.4.1 Specific comments ...... 199 11.4.2 General comments ...... 199 11.5. Main challenges of INS in the area of infrastructure ...... 200 11.5.1. Infrastructure development ...... 200 11.5.2. Public acceptance ...... 200 11.5.3. Human resources ...... 201 11.5.4. International cooperation ...... 201 11.6. Results of the INS evaluation in the area of infrastructure ...... 202
CHAPTER 12. R&D AND COLLABORATIONS FOR INNOVATIVE CNFC-FR ...... 208 12.1. Introduction ...... 208 12.2. Economics ...... 208 12.3. Safety and environmental impact ...... 211 12.4. Economy and safety of fuel cycle facilities ...... 212 12.5. International collaboration ...... 212
REFERENCES ...... 214
ANNEX A: OUTLINE OF A FRENCH CASE STUDY ...... 218
ANNEX B: JAPAN’S EVALUATION RESULTS OF THE INPRO JOINT STUDY ...... 228
ABBREVIATIONS ...... 290
CONTRIBUTORS TO THE JOINT STUDY ...... 294
CONTRIBUTORS TO DRAFTING AND REVIEW ...... 296
SUMMARY
An assessment study (called Joint Study) was started in 2005 and completed in 2007. Canada, China, France, India, Japan, the Republic of Korea, the Russian Federation, and Ukraine participated. The objectives of the study were to: • Assess the potential of a nuclear energy system based on a closed fuel cycle (CNFC) with fast reactors (FR) regarding its sustainability using the INPRO methodology; • Determine milestones for the CNFC–FR system deployment; and • Establish frameworks for, and areas of collaborative R&D work. It was agreed to use for the assessment as a reference system a commercial CNFC–FR system, deployable in the near term (20 to 30 years) based on proven technologies, such as sodium coolant, mixed oxide (MOX) pellet fuel, and advanced aqueous reprocessing technology. Results of the nuclear energy system assessment using the INPRO methodology The INPRO methodology requires an assessment in seven areas to confirm the sustainability of a nuclear energy system: economics, infrastructure, proliferation resistance, physical protection 1 , environment (impact by stressors and availability of resources), waste management and safety. The main results of the multidimensional assessment of the CNFC– FR system could be summarized as follows: • Availability of resources: Recycling of plutonium (together with uranium) in spent fuel of CNFC–FR systems leads to practically inexhaustible resources of fissile material (and fertile material), i.e. such a system might de facto be considered as a renewable energy source. Globally, there is sufficient spent fuel available for reprocessing Pu to be used as fuel for FR. However, in some countries with an expected high growth rate of their national nuclear power program lack of spent fuel as a resource of Pu may impede an optimal deployment of CNFC–FR systems. Thus, the Joint Study concluded that a CNFC–FR system is well suited for and might require a regional or multilateral approach as no individual country participating in the Joint Study reflects the full set of factors that favour development and deployment of such a system. Examples of important favourable factors are predicted high growth of energy demand and large resources of Pu available in spent fuel. • Impact of stressors: CNFC–FR systems avoiding mining/enrichment steps in their fuel cycle show a significantly reduced environmental impact caused by a much lower release of non-radioactive elements compared to current licensed thermal reactor systems. Additionally, the radiation dose of CNFC–FR systems on the public is demonstrated to be far below regulatory limits. • Waste management: The CNFC–FR system meets all INPRO requirements of an effective and efficient nuclear waste management. By recycling of specific (heat producing and long lived) nuclear fission products and minor actinides in addition to plutonium (together with uranium), the CNFC–FR system has the potential to significantly reduce the heat load, mass/volume and radiotoxicity of high level waste to be deposited. The reduction of heat load enables to store more waste per volume of
1 The area of physical protection of the INPRO methodology was not covered in the Joint Study as the description of the assessment method for this area was not yet available during the performance of the Joint Study.
1 rock, and the removal of actinides and specific fission products from the waste decreases the time required to manage nuclear high level waste from a geological time scale (several 100 000 years) to a civilization time scale (several 100 years). However, in comparison to a once through fuel cycle (OTFC) reprocessing of spent fuel in a CNFC produces several additional secondary nuclear waste forms (e.g., losses in the processes) most of them needing geological disposal. • Safety: Safety characteristics of the CNFC–FR system meet the current safety standards. A comparison of a CNFC–FR with a thermal reactor system showed that disadvantages of the fast neutron system were compensated by its inherent safety features and additional engineered safety measures. A probabilistic analysis performed for the Russian BN-800 fast reactor design confirmed that its innovative design features lead to a significant reduced risk of severe accidents, thus relieving the need for relocation or evacuation measures outside the plant site. • Proliferation resistance: CNFC–FR systems show several features that result in comparable or higher proliferation resistance compared to thermal reactors with a once- through fuel cycle (OTFC). The higher proliferation resistance of CNFC–FR is justified by eliminating enrichment of uranium and avoiding the accumulation of Pu in spent fuel (“plutonium mines”) in an OTFC, excluding Pu separation in advanced reprocessing technologies, and by the possibility to produce fresh fuel with a high radiation level and to reduce fuel transportation via collocation of FR and fuel cycle facilities applying pyro-processing technology. The use of a higher fissile content in the FR fuel results in a decrease of proliferation resistance in comparison to thermal reactor systems. • Infrastructure: The INPRO basic principle in the area of infrastructure asks for the availability of regional and international arrangements to limit the necessary effort to establish the necessary infrastructure for a nuclear energy system. As stated above in the area of availability of resources, the Joint Study concluded, a CNFC–FR system is well suited for and might require such new regional or international arrangements as it is capable of converting spent fuel of all reactors into a valuable energy resource, thereby offering the opportunity expanding fuel cycle front end and backend services on a multinational basis to technology holder as well as to technology user countries. Looking at the national legal infrastructure of the countries participating in the Joint Study it was found that the legal frame work needed to operate a nuclear energy system is well established and is deemed sufficient to cover also future CNFC–FR systems. However, regional or international approaches might require new international legal infrastructure. The industrial infrastructure and human resources to design, manufacture, construct and operate a CNFC–FR system are available in most countries participating in the Joint Study. • Economics: The designs of currently operating fast reactors with a closed fuel cycle are not completely economically competitive against thermal reactor systems or fossil power systems due to high capital costs. However, the necessary modifications of the design, such as simplifying the design, increasing the fuel burnup, constructing small series are integrated into the development programs of all Joint Study participants. These modifications will make electricity costs produced by CNFC–FR systems comparable to those of thermal reactor and fossil fuelled power plants.
2 Milestones of installment of CNFC-FR systems
China plans to increase its nuclear capacity significantly from about 7 GW(e) to 60 GW(e) by 2030 installing mainly PWR of different sizes. Starting around 2020, fast breeder reactors are planned to be added and these will become the dominant type of nuclear reactors by the end of the 21st century. The cores of the fast reactors are designed with moderate breeding rates ~1.1 to 1.2 at the beginning and could be replaced later by core designs with high breeding rates such as ~ 1.5. In France, for the purpose of the study, nuclear power capacity of about 63 GW(e) is assumed to remain constant throughout the 21st century. The existing fleet of PWR is to be replaced by about 2030 with EPR (Generation III+) reactors. Thereafter fast reactors to be operated as Pu burners (with a conversion rate about 1) should become the dominant option of nuclear energy supply. India plans to build primarily PHWR and PWR to satisfy the predicted large increase of capacity of nuclear power from 3 GW(e) to about 30 GW(e) by 2020. Starting in 2010 a fleet of fast breeder reactors with – similar to China – a moderate breeding ratio of ~ 1.1 to 1.2 at the beginning is planned to be installed, and later changed to core designs with a high breeding ratio of ~ 1.5. Finally, advanced PHWR – using an advanced thorium fuel cycle – are foreseen to be installed leading to a total nuclear capacity of 275 GW(e) by 2050. Japan – similar to France – expects a moderate increase of nuclear generation capacity from 47 GW(e) to 60 GW(e) by 2030. Thereafter it is foreseen to keep the nuclear generation capacity constant till the end of the 21st century. Starting around 2050 fast breeder reactors (with a low breeding ratio of ~ 1.03 to 1.10) could gradually replace LWR. The Republic of Korea investigates several options for satisfying the predicted increased capacity of nuclear power from about 17 GW(e) to 27 GW(e) by 2015. One option foresees the installment of additional water cooled reactors only, and the two other options include the installment of fast reactors to be operated as Pu burners (with a conversion rate about 1) after about 2030. In the Russian Federation the nuclear capacity should increase significantly from about 22 GW(e) to 81 GW(e) by 2050. The existing fleet of VVER is to be replaced by Generation III reactors of type AES2006 and around 2030 fast breeder reactors are foreseen to replace gradually the thermal reactors. Breeding ratios are expected to be within a wide range from about 1.0 to 1.6 at different stages of the nuclear power program. Looking at the development programs of a CNFC–FR system in the countries participating in the Joint Study it is to be noted that: • All countries developing fast reactor technology selected a stepwise introduction of the technology, starting with a small experimental reactor (< 50 MW(th)) to test the feasibility of the concept, then installing a prototype (several 100 MW(th)) to confirm all technical issues are resolved, thereafter constructing a first commercial size reactor (several 1000 MW(th)) to proof its competitiveness, and finally install a series of commercial reactors by 2020 to 2050. A similar stepwise approach is applied for the development of the associated fuel cycle technology. • The time schedule of installment of the CNFC–FR system strongly depends on the global as well as on the national development of nuclear power. The higher the growth rate of nuclear power capacity either assumed globally or planned in the country, the earlier the installation of a CNFC–FR system is required to assure the availability of
3 cheap fissile material. Differences between countries in the installation schedule are mainly caused by different predicted growth rate of the national nuclear power program. • Countries with a large nuclear power program established for a long time and with moderate or no planned increase of their nuclear power capacity, i.e. France, Japan, or Korea, expect to accumulate enough Pu in spent fuel needed for a fast breeder program, and are therefore focusing on the development of core designs of FR with low to moderate breeding rates, e.g., Pu burners.
• Countries like China, India and the Russian Federation with rather limited contribution of nuclear power to their total energy supply but with a planned large and rapid increase of their nuclear power capacity expect not to accumulate enough Pu in spent fuel for their planned FR program, and therefore aim initially already at core designs with moderate breeding ratios and consider in the long term core designs with breeding rates as high as possible to avoid a shortage of fissile material.
R&D defined in the Joint Study The Joint Study concluded that a comprehensive program of R&D is absolutely essential in a variety of areas (especially, for economics and safety) with an inter-disciplinary approach and international collaborations wherever possible to make a CNFC–FR system a viable alternative to conventional sources of power. As capital costs of currently operating (sodium cooled) FR were 40 % up to three times higher than capital costs of thermal reactors, several possibilities for reduction of capital costs were presented. For the improvement of FR safety, R&D is needed to develop efficient and cost-effective shielding materials such as boride/rare earth combinations, and achieve (radiation) source reduction by adequate measures such as use of materials which do not get activated. In the Joint Study the following INPRO collaborative projects related to fast reactors have been proposed and are currently underway: • A Global Architecture of nuclear energy systems based on thermal and fast reactors including a closed fuel cycle (called GAINS).
• Integrated Approach for the design of safety grade decay heat removal system for liquid metal cooled reactor (called DHR);
• Assessment of advanced and innovative nuclear fuel cycles within large scale nuclear energy system based on CNFC concept to satisfy principles of sustainability in the 21st century (called FINITE);
• Investigation of technological challenges related to the removal of heat by liquid metal and molten salt coolants from reactor cores operating at high temperatures (called COOL);
Feedback from the Joint Study on the INPRO methodology Besides detailed proposals how to improve the INPRO methodology in specific areas several general proposals have been made to improve the INPRO methodology as set out below. The INPRO methodology should be extended to enable a clearer distinction (discrimination) between different options of nuclear energy system components under development.
4 An approach should be developed how to treat different level of uncertainty associated with stages of development. In particular for the INPRO area of environment and proliferation resistance a need for further development of the assessment approach was expressed.
5 CHAPTER 1 OVERVIEW OF THE ASSESSMENT RESULTS
Studies on the capability of different innovative nuclear energy systems (INS) to provide adequate answers to challenges of a large scale nuclear power deployment are an integral part of the INPRO project initiated by a group of IAEA Member States in the year 2000 [1]. INPRO methodology [2], consisting of a set of basic principles, user requirements and criteria, was developed for the area of economics, safety, environment, waste management, proliferation resistance, physical protection and infrastructure with a goal to evaluate the potential of an INS to contribute to sustainable energy supply. Several participants of INPRO (Canada, China, France, India, Japan, Republic of Korea, Russian Federation, and Ukraine) agreed to assess jointly the capability of an INS based on a closed nuclear fuel cycle with fast reactors (CNFC–FR) to comply with the objective of sustainable nuclear energy development, by applying the INPRO methodology described in the INPRO manual [3]. The main results of this study regarding the capability of an INS to meet the requirements of the INPRO methodology and enhance sustainability features of nuclear power are presented in this chapter.
1.1. SCOPE OF DESIGNS AND TECHNOLOGIES UNDER CONSIDERATION
Fast reactor and closed fuel cycle technologies have been under development since the beginning of the nuclear power era. At those times, the potential capability of such a fast reactor system to efficiently extract energy from uranium resources (60 times more efficient in comparison with thermal reactors operating in an open cycle) was considered as a main argument for the development of fast reactor technology. By the middle of the eighties, the feasibility of all basic technologies for a fast reactor system had been demonstrated at demo and industrial levels, including the mastering of sodium technology, MOX (mixed oxide uranium-plutonium) fuel fabrication technology, and aqueous PUREX reprocessing technology for the treatment of (thermal and) fast reactor fuels. Nevertheless, development of this reactor system was practically frozen in the nineties due to several reasons: The absence of near term plans for construction of new plants in nuclear power leading countries and uncertain global nuclear power prospects for coming decades were the most important reasons for stagnation. At the beginning of this century, expectations for a new phase of nuclear power development raised again the interest on fast reactor and closed fuel cycle technologies, particularly in the countries with large scale nuclear power programs. Japan scheduled to restart its prototype sodium cooled fast reactor MONJU in 2009. China planned to finish construction of the Chinese experimental fast reactor CEFR by 2008. In India, the prototype fast breeder reactor (PFBR) was scheduled to be commissioned by 2010. India is also undertaking the construction of a co-located fast reactor fuel cycle facility. Beyond PFBR, India plans to build four reactors of a similar type by 2020. The Russian Federation has restarted construction of its sodium cooled BN-800 with MOX fuel; commissioning of this reactor is scheduled in 2012. China, France, India, Japan, Republic of Korea and Russian Federation are carrying out (with different emphasis) R&D aiming to investigate the potentials of various types of reactor fuel, coolant, fuel fabrication and reprocessing technologies to achieve improvements in the area of economics, safety, waste management, resource sustainability, and proliferation resistance. Feasibility of most of these innovative ideas is still to be demonstrated.
6 For this study it was agreed to use as a reference a near term innovative nuclear energy system based on a closed fuel cycle with fast reactors (INS CNFC–FR) to rely only on proven technologies – sodium coolant as coolant, MOX pellet fuel and aqueous reprocessing. The potential of novel ideas is also discussed in the study but with a lower level of confidence.
1.2. MULTIDIMENSIONAL ASSESSMENT OF THE INS
1.2.1. Safety
Current safety status Safety performance of existing fast reactors and their fuel cycle is founded on the international basic nuclear safety philosophy. Some physical and technological properties of the fast reactor system facilitate the realization of safety principles, while others impede it. For example, sodium leakages in demo fast reactors caused certain concerns regarding chemical activity of sodium coolants and created a necessity of further developmental activities, such as procedures, equipment and training sufficient to prevent/manage sodium “fires”. A few events of small reactivity variations occurred in operating sodium cooled fast reactors. However, they did not lead to abnormal situations or failures but caused a reactor shutdown and called for detailed investigations. These incidents also confirmed the need for further international harmonisation of national regulatory requirements to provide adequate responses to these events, including development of a common approach to specific fast reactor safety issues such as the sodium void effect, sodium leakages, etc. On the whole, experience gained has demonstrated that sodium cooled fast reactors and their fuel cycle are able to meet current safety standards, provide a reliable and flexible operation with low levels of unplanned outages, a high overall plant availability and a high annual load factor (up to 80 %). Proven operation of main equipment without major repair reached 200−300 thousands of hours. Safety features of near term CNFC-FR Enhancing of the defense-in-depth in sodium cooled near term fast reactor systems 2 through optimized design options/configurations of active and passive safety systems makes it possible to mitigate the risk of severe accidents by at least an order of magnitude as compared to existing designs. The calculated core damage frequency is well below 10-5 per reactor-year for near term CNFC-FR, which is the value defined by the IAEA as a threshold for an adequate safety level. The cumulative frequency of a large release of radioactivity by a near term fast reactor to public domain was determined to be below 10-7 per reactor-year; that is an order of magnitude less than the value for existing designs. Safety analysis of fast reactors demonstrates that the reduction in occupational and public exposure prevents a need for relocation or evacuation measures outside the plant site in case of a major release of radioactivity. Superior safety characteristics for future fast reactor systems To achieve enhanced safety characteristics in future innovative CNFC−FR systems 3 , designers place increased emphasis on inherently safe characteristics and passive systems. A major release of radioactivity to public domain has to be practically excluded in the design. Intensive research, development, and demonstration (RD&D) has to be carried out to prove
2 Near term systems are systems available for deployment within the next 10 to 20 years. 3 Future innovative systems should be available for deployment in around 20 to 30 years.
7 safety advantages of fast neutron systems with emerging design concepts, e.g., advanced coolants such as lead-bismuth, lead, or gas, advanced fuels such as metal, nitride, and carbide, and/or advanced physical concepts in case of dedicated plutonium burners. Technical feasibility, reliability, and improved economics of such advanced systems have also to be demonstrated. Current safety standards are met by a few operating demonstration and demo/commercial INS CNFC–FR (with sodium cooling) while more extensive confirmation is still needed. The near term target for the INS CNFC–FR is to achieve a risk level of severe accidents at least an order of magnitude less than for existing designs. Enhanced safety with inherent safety features and passive systems is being developed for INS CNFC–FR systems, and has to be demonstrated comprehensively in regard to technical feasibility, reliability, and improved economics.
1.2.2. Environment and resource sustainability
Low adverse environmental effects Specific features of the CNFC–FR assure minimal adverse effects on the environment and humans during operation. Releases such as gas-aerosol, 131I, and long- and short-lived nuclides are below the threshold of recording instruments; inert radioactive gas releases are mainly caused by experimental studies in hot cells. Consequently, radiological impact of a CNFC-FR is far below regulatory admissible levels. Even though heat rejection from a fast reactor is essentially less than from a thermal reactor, further reduction should be attempted to reach comparability with matured fossil fuelled power plants. Experience gained testifies that fuel recycling and avoiding mining/enrichment steps in a CNFC-FR can reduce by more than three times the emission of greenhouse gases (GHG) (CO2, CH4, N2O) as well as emission of air polluting gases (NOX, SOX) and harmful particles. Resource sustainability The CNFC–FR provides a possibility to exploit the vast resources of natural 238U for energy conversion. A CNFC–FR system leads to a net increase of the nuclear energy resource base by more than 60 times in comparison to present nuclear energy systems with thermal reactors operating in an open fuel cycle mode. Based on practically inexhaustible natural resources of 238U the CNFC–FR can be de facto considered as a renewable energy resource suitable for a large scale global exploitation. It is expected that only by use of fast reactor systems, nuclear power will have the potential to become a significant player in the energy mix needed for global sustainable development. The Joint Study identified as an important task for further environmental studies a comprehensive evaluation of ways decreasing the use of fresh water and other valuable natural resources in a closed fuel cycle, specifically at the stage of reprocessing. Innovative CNFC-FR systems
The current fast reactor technology – with a plutonium breeding ratio (BR) in the range between l.0 and 1.3 – permits introducing efficiently 238U in the energy balance through recycling of plutonium accumulated from thermal reactors. For those countries that have large scale thermal reactor fleets at present and do not plan for a significant addition of nuclear power capacity in coming decades, the current available level of breeding might be quite adequate. Other countries like China, India, and, to a certain degree, the Russian Federation that do not have sufficient plutonium for large fast reactor programs, are developing advanced fuels (nitride, metal) with an objective to increase the breeding ratio to 1.4–1.5. This would
8 provide impetus for an accelerated deployment of fast reactors. Fast reactors with a high breeding ratio might also be used in the future for generation of denatured uranium (238U +233U) fuel for thermal reactors; fissile 233U could be generated in thorium blankets of fast reactors. The environmental effects of the demonstration and near term INS CNFC–FR are well within the performance envelope of current nuclear energy systems delivering similar energy products with the lowest GHG emissions among them. The INS CNFC–FR has a break- through potential to meet the sustainability requirements related to the resource area. The INS CNFC–FR is suitable for a large scale national and global exploitation due to long term availability of natural resources. Realization of this potential for a nuclear power system, as a whole, would depend on the share of fast reactors in the whole energy system. Development of advanced fuels might help to accelerate, in some countries, deployment of fast reactors and/or generate denatured uranium fuel for thermal reactors, if needed.
1.2.3. Waste management
Waste minimization Minimization of radioactive waste by recycling and reuse of fissile materials that would otherwise be high level (alpha-emitting) radioactive waste (HLW) containing long-lived toxic components was an additional argument in favour of CNFC–FR development at the very early stage of nuclear power development. Currently, technical feasibility of MOX fuel recycling and vitrification of HLW arising from reprocessing of reactor fuel has been demonstrated on an industrial scale. By multi-recycling of fuel in fast reactors, 90 % mass reduction of fuel waste requiring repository disposal can be achieved, which means a 5–7 times reduction in the number of repositories required. The objective of waste minimization is thus a key driver for introduction of an INS CNFC–FR. Waste optimization Having a clear ultimate perspective in HLW management, the technology of CNFC-FR provides flexibility by optimal use of resources (funding, space, capacity, etc.) for achieving a balance between long-term perspectives and short-term risks and costs. For a group of countries with a sizable fleet of light water cooled reactors (LWR), e.g., France, Japan, and the Russian Federation a CNFC-FR system is considered as an option to avoid the accumulation of plutonium in spent fuel from LWR and to decrease the quantity of spent fuel. Development of novel waste management technologies Recycling of MOX fuel and safe conditioning of residual waste (containing fission products and minor actinides) after reprocessing demonstrated in France will remain a practical option for coming decades. Nevertheless, innovative technologies are under investigation aiming to reduce further the burden on repositories by optimization of fissile material management and transmutation of minor actinides in fast reactors or accelerator driven systems (ADS). The feasibility of these technologies still has to be demonstrated. The contribution of transuranic elements (TRU) to occupational and public exposure in different options of nuclear fuel recycling also has to be better understood. Generally, a technological approach to waste optimization based on recycling and reuse of radioactive materials has a high potential, not only in management of high level waste (HLW) from nuclear fuel but also in treatment of radioactive wastes arising from decommissioning of nuclear installations both of once through (OTFC) and closed (CNFC) nuclear fuel cycles.
9 Avoidance of undue burden on future generations It is judged that the INS CNFC–FR has a break-through potential to meet the fundamental principle of sustainability: to use nuclear fuel resources effectively and not to impose undue burden on future generations. Recycling of Pu and minor actinides in the future within an INS would decrease the level of radiotoxicity of waste by more than two folds of the acceptable level today. Eventually, radioactive waste from an INS with a CNFC could be managed in such a way that its end state reaches radioactive equivalence4 with uranium ore thousand times earlier than in case of an OTFC. In this case, a long term radioactivity risk on future generations is practically excluded since integrity of radioactive barriers of the waste repositories can be guaranteed for the time period needed for the waste reaching radioactive equivalence with uranium ore. Safe conditioning of waste arising from plutonium recycling is industrial reality today and was an important practical milestone in reaching ultimate goals of a closed fuel cycle strategy. The INS CNFC–FR has practically demonstrated its potential to meet all present day requirements related to waste management.
1.2.4. Proliferation resistance
For the INS CNFC–FR the issue of assuring proliferation resistance (PR) is a matter of principal concern, since the argument of a high proliferation potential of fast reactors has adversely affected development of the technology in the past. In accordance with the INPRO methodology, PR was assessed for INS CNFC–FR, taking into account intrinsic features that result from technical design of the nuclear energy system, and/or extrinsic measures based on States’ decisions and undertakings. Potential to enhance PR intrinsic features Different engineering, design, and technological proposals aiming to enhance intrinsic PR features of the INS CNFC-FR are extensively discussed and examined in the study. Examples are: the highly proliferation resistant end-state of the closed fuel cycle (avoiding of Pu “mines”); exclusion of Pu separation in reprocessing; no need for U enrichment processes; possibility to seal a reactor core for 10–20 years of fuel campaigns; limiting the length of fuel transport by co-location of fuel cycle facilities and reactors. It is judged that implication of intrinsic features in the INS CNFC–FR could reduce the risk of misuse of facilities of the INS to a level equal or lower than the risk of misuse of facilities for thermal reactors with OTFC. Limitations of technical approaches to PR Though the potential of the INS CNFC–FR is high for enhancing PR by implementation of intrinsic features, technical barriers alone could not give an irrefragable answer to all proliferation threats. In absence of safeguards, no technical solution has been found with a potential to block in principle the possibility of misuse of enrichment and reprocessing facilities. Technical solutions may impede misuse of a reactor or facility, but they would not be able to prevent misuse of a complete nuclear system, or avoid misuse of knowledge or a break-out scenario. Besides, a complicated technical fix may have negative implications on INS’ economic and safety features. Therefore, both extrinsic measures and intrinsic features are essential and they should go hand in hand, since neither is sufficient by itself.
4 Radioactive equivalence of waste with uranium ore means that the waste has the same radiotoxicity as uranium ore.
10 Role of extrinsic measures Political agreement in the area of non-proliferation settled by international treaties and commitments of States is the fundamental basis of all extrinsic measures both for the open and closed fuel cycle systems. Responsible governments must control, to the extent possible, the know-how relevant to produce and process either highly enriched uranium (enrichment technologies) or plutonium (reprocessing technologies). To support the non-proliferation regime, activities of the IAEA and other international institutes/agencies should be intensified. New technologies in safeguards, e.g., for monitoring and inspections should be developed and applied. Countries with developed nuclear power and related infrastructure can provide assistance for newcomers, thus contributing to strengthening the non-proliferation regime. Deployment of multinational fuel cycle centers may help coping with all listed problems to a large extent.
1.2.5. Infrastructure
Legal infrastructure and public acceptance Legal framework including a nuclear law in the countries participating in the Joint Study is established in accordance with international standards, and international nuclear organizations play an important part in further improving the legal basis. Long-term policy and commitment by governments to support sustained deployment of INS CNFC–FR generally exists but political changes remain always possible. Public acceptance is a critical point in ensuring a negligible risk for changing nuclear policy. The worldwide safe operation of nuclear installations along with relevant and understandable information to public and media, involvement of the public in decision making processes and other measures providing efficient interaction with society should eventually result in enhancing a positive public attitude towards further development of nuclear power as a way to assure sustainable energy supply and meeting global environmental challenges. It was judged that information provided to the public regarding INS CNFC–FR is sufficient according to best international practice. Professional perspectives Insufficient availability of young scientists and engineers to the domain of nuclear science and technology in many countries is one of the most serious negative consequences of nuclear power stagnation. It is hoped that nuclear renaissance will eventually change public and governmental opinion attracting the young generation to nuclear science and industry. This should provide a long-term perspective for technology development, especially development of innovative systems like the INS CNFC–FR. Multinational approach to industrial and economic infrastructure A survey performed in the Joint Study showed that national conditions are complementary, which creates promising prerequisites for mutually beneficial long term collaboration. No single country, taken separately, meets the full set of requirements favouring development of CNFC–FR − high energy demand, high level of nuclear technology and infrastructure maturity, high resource of nuclear fuel, etc. − while the global nuclear energy system does. Assurance of nuclear fuel supply and services is one of the most prospective areas for collaboration. Plutonium in spent fuel of thermal reactors is being destined to be disposed or burned in one part of the world while in another part (China and India) the lack of the material as a fuel for fast reactors threatens to restrict a high rate of nuclear power deployment. International legal and institutional infrastructure has to be significantly advanced to realise the potential of mutually beneficial collaboration. Efforts are made to widen the multilateral
11 approach to nuclear fuel cycles (MNFC) under strong IAEA safeguards. The INS CNFC–FR − being capable of converting spent fuel of thermal reactors into a valuable energy resource − provides a principal opportunity for a significant expansion of fuel cycle front and back end services on a multilateral basis. Transition to a global nuclear infrastructure could evolve realizing the potential of synergy between fast and thermal reactors as well as the synergy between countries as a qualitatively new stage of nuclear power development. This will provide new perspectives of growth for matured nuclear industries, facilitate the use of nuclear power for newcomers, and mitigate the consequences of a nuclear phase out in the case of radical change of energy policy in some countries.
1.2.6. Economics
Economics, even though strongly country-dependent, is key for a large-scale deployment of INS CNFC–FR. Cost Technical feasibility and cost effectiveness of an INS based on CNFC–FR has been demonstrated in technology-leading countries. Fuel burn-up in excess of 130 GWd/t of heavy metal was achieved in several fast reactors. Two fast reactors are currently operating with a thermal efficiency of 43–45 %; that is the highest value achieved in nuclear power operation [4]. Yet, the current CNFC-FR does not assure meeting the INPRO economic acceptance limits. Knowledge accumulated so far indicates that, in a time span of 10–20 years, the cost improvements for sodium cooled fast reactors would be focused on design simplification, increasing fuel burnup, etc., ultimately, to achieve cost comparability with thermal reactors and steam turbine fossil-fueled power plants. Reaching competitiveness with fossil-fueled combined cycle turbine plants also seems to be a feasible task under the condition that extensive RD&D towards commercialisation of fast reactor would be carried out. Besides advanced CNFC–FR with MOX fuel, other fast neutron systems based on heavy metal and gas coolants, advanced fuels and technological innovations in CNFC may help to improve significantly economic competitiveness after establishing technological feasibility. Projected rise of fossil fuel costs confirmed by current trends are in favour of long-term competitiveness for both thermal and fast reactor systems. Carbon emission credits could give additional incentives for nuclear power. Ability to finance and investment risk Providing total investments required for the introduction of an INS CNFC–FR, including establishment of infrastructure, carrying out of large-scale RD&D and construction of necessary facilities and installations, is a crucial point for making this energy option available. It is shown in the Joint Study that up-front costs of sodium cooled fast reactors that drive the cost structure of fast reactors can be reduced by learning-by-searching, which necessitates carrying out extensive R&D continuously. At the same time, it is expected that construction of sodium cooled fast reactors in some countries participating in the Joint Study will result in reducing the costs due to learning-by-doing and building in series. However, in any case governmental support would be inevitable during the introductory phase of an INS CNFC–FR system. Decreasing necessary total investments will accordingly minimise the investment risk and provide a return on investment, comparable with competing energy technologies. Nevertheless, only countries with sizeable nuclear power programmes will have enough
12 incentives to invest high funds and meet high risks. That’s why, most probably, the first phase of the system commercialization will be realized in the confined group of these countries.
First of a kind INS CNFC–FR systems did not meet confidently the INPRO economic requirements. Commercial INS with a series of fast reactors based on available technologies could be affordable in 10–20 years in the countries mastering the technology. The INS as a part of a global nuclear system will be available and affordable via further improvement of economics based on identified RD&D and expected industrial deployment.
1.3. OVERALL JUDGMENT
Assessment of current INS CNFC–FR Some first of a kind (FOAK) systems based on CNFC–FR with sodium as coolant demonstrate operational characteristics that meet the requirements of current regulatory basis. Compliance with economic indicators of INPRO methodology has not been confidently demonstrated by the systems to date. Near/medium term expectations In the diagram presented in Fig. 1.1 assigning of a number higher than ‘one’ for the INS CNFC–FR compared to a NS OTFC−TR (nuclear energy system with thermal reactors and once through fuel cycle) indicates ability of the INS in the near/medium term prospect to enhance sustainability features of the current nuclear system and provide it with a new quality.
ECONOMICS
2 INFRASTRUCTURE SAFETY
OTFC-TR
PROLIFERATION RESISTENCE CNFC-FR WASTE MANAGEMENT
FIG. 1.1. Multidimensional assessment of an operating nuclear energy system OTFC−TR and a near/medium term CNFC−FR (the higher number on the diagram − the better the performance). Reaching confidential economic competitiveness with the NS OTFC−TR remains an important goal of the near-term agenda for the CNFC system with sodium cooled fast reactors. Even though close to commercialization, the CNFC-FR with sodium cooling still needs execution of sizable RD&D to demonstrate the success of innovative features. In countries developing CNFC−FR, system introduction is supported by funding from governmental and industry sources. It is expected that improvements through identified RD&D together with series construction of sodium cooled fast reactors would result in cost competitiveness of the near term INS CNFC−FR with sodium cooling with the NS OTFC-TR (Fig. 1.1).
13 Another area of the CNFC−FR with sodium cooling that needs significant improvement to realize the potential of the system is infrastructure, especially the aspects of long term government commitment and public acceptance. There is potential to improve these aspects due to availability of favorable waste management strategies, i.e. lower production of waste and better transmutation capabilities in CNFC−FR systems, as well as due to abilities of CNFC−FR systems to install multinational arrangements of nuclear fuel cycles. One of the major conclusions of the Joint Study is an appeal to improvement of national and global nuclear infrastructure applying both technological and institutional innovations. The capability of the near/medium INS CNFC−FR to comply with INPRO criteria of sustainable energy supply in other areas of INPRO methodology, apart from economy and infrastructure, should be better or much better, compared to current nuclear systems (Fig. 1.1). Most probably, the gradual introduction of the near/medium term INS CNFC−FR will happen in a time horizon of 20−40 years in the countries that will continue mastering the technology with the objective to enhance sustainability of their nuclear power system. Though it may be only a few countries, their population and use of energy can be a significant share of the world. Long term perspectives Assessing the long term perspectives of the INS CNFC−FR, emphasis is given to the potential of technological and institutional innovations. Technological progress through intensive R&D at national and, increasingly, international levels, will undoubtedly remain the main driving force to make the INS CNFC−FR with different types of fast reactors available and affordable. It is hoped that after commercialization in the lead-countries, the INS CNFC−FR will contribute to assuring front and back end nuclear fuel services worldwide and thus become a mechanism to provide a global nuclear response to energy challenges in the 21st century. The IAEA INPRO project seems to be the right forum for development of a multinational vision of a transition from the present nuclear systems to a future global nuclear system with enhanced sustainability features. Improvements through R&D activities and construction in small series are expected achieving full compliance by a medium term INS CNFC−FR with the criteria of sustainability defined by the INPRO methodology. A scaled introduction of INS would increase sustainability features of nuclear power and therefore it remains an important aim for technology development.
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CHAPTER 2 OUTLINE OF THE JOINT STUDY
2.1. POSITION OF THE JOINT STUDY IN INPRO
The Joint Study on an innovative nuclear energy system (INS) based on a closed nuclear fuel cycle with fast reactors (CNFC−FR) is part of the IAEA international project on innovative reactors and fuel cycle (INPRO) [1]. INPRO was launched in 2001, based on resolution of the IAEA General Conference GC(44)/RES/21, 2000. One of the main goals of INPRO is to help to ensure that nuclear energy is available in the 21st century in a sustainable manner. INPRO methodology, consisting of a set of basic principles, user requirements and criteria, was developed in phase 1A of the project in the domain of economics, safety, environment, waste management, proliferation resistance, physical protection and infrastructure [5]. Manuals were prepared in the respective domains [3]. Validation of the methodology was performed in Phase 1B of the project by performing various case studies, which helped to refine and adjust the methodology [2]. The INS CNFC−FR has many attractive features for providing sustainable energy supply in the 21st century. Since available uranium and thorium resources can be utilized effectively, CNFC−FR ensures a large resource suitable for large scale exploitation. Fast reactors can be used to burn minor actinides, can provide liquid metal technology and high temperature design inputs for accelerator driven systems (ADS), fusion and high temperature reactors (HTR), and can generate electricity at competitive cost over long periods. In order to evaluate jointly these (and other) advantages as well as associated challenges, the Joint Study assessing an innovative nuclear energy system based on CNFC−FR was initiated by the Russian Federation in October 2004 at a meeting of countries participating in INPRO. Member States having more than half of the world population and large users of energy in coming decades are affiliated to the Joint Study (Fig. 2.1). In the initial stage, China, France, India, Republic of Korea and the Russian Federation joined as members, and Japan joined as observer in December 2004; in 2005 Ukraine joined; Canada and Japan became participants in 2006.
FIG. 2.1. Countries participating in the Joint Study. (Russia = Russian Federation, Korea = Republic of Korea)
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The Joint Study has enabled consolidation of the experiences of various countries, thereby creating a possibility of achieving a significant enhanced share of nuclear energy in coming years by learning from experiences. The study enables sharing of possible innovative features and R&D outputs generated by participating countries. Further, the study provides a holistic and global perspective and establishes enhanced synergy with relevant international projects by providing a frame work and platform for collaboration.
2.2. MAIN OBJECTIVES
The assessment of INS was carried out in accordance with the overall goals of INPRO and guiding documents of the Joint Study developed and approved by the participating parties. The main objectives of the Joint Study were (Fig. 2.2):Consolidation of perspectives of participating countries on the future of CNFC−FR and its role in nuclear programmes, exchange of experiences, identification of the optimum structure of the system, and milestones for its deployment at a national and global level; • Identification of a reference CNFC−FR for an INPRO assessment; • Assessment of the INS CNFC−FR using the INPRO methodology requirements for sustainable development in the area of economics, environment, waste management, safety, proliferation resistance, infrastructure, and providing a feedback towards improvement of the INPRO methodology; • Identifying areas of collaborative R&D work; • Recommendations for enhancing co-operation in areas of mutual interest.
Methodology INPRO improvement Methodology
INS data Assessment Joint Study Country/Region/ World Common R&D context data identification
Multilateral CRPs programmes Others
FIG. 2.2. Scope of the Joint Study (CRP=IAEA coordinated research projects).
2.3. ORGANIZATIONAL FEATURES AND MANAGEMENT OF THE PROJECT
The Joint Study was open to all IAEA Member States to participate in, provided they accepted the terms of reference agreed upon when the Joint Study was set up by the parties. Each party in the Joint Study funded its work by means of national contributions. Each party had the right to assign any of its participants to work on the Joint Study and was responsible for their efficient performance with respect to the work agreed to under the Joint Study, and
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for compliance with the rules regarding confidentiality and exchange of information agreed upon by the parties. No exchange of technologies and know-how was anticipated. Any activity undertaken as part of the Joint Study complied with the Statute and operating principles of the IAEA. For the purpose of scientific and administrative management of the Joint Study, the participating parties established a scientific and technical committee (STC). The meetings of the STC (see Fig. 2.3) were held alternately at the IAEA or in a country participating in the Joint Study. National working groups were established for each INPRO methodology area (economics, safety, environmental protection, waste management, proliferation resistance, and infrastructure). As a part of an IAEA project, the Joint Study profited from the collaboration with specialists from the IAEA.
FIG. 2.3. Participants of the 5-th Joint Study STC meeting in Cadarache, France. In the terms of reference (TOR) of the Joint Study it was defined that the study will be carried out in two main stages. At the first (preparatory) stage scenarios of nuclear power development worldwide and in participating countries, as well as INS concepts based on a CNFC−FR were identified. Several INSs were defined for review. The appropriate parameters for analysis of INSs were elaborated and agreed upon jointly. The scale of the development of the power generation sector up to the year 2050 as a minimum case was specified as an input parameter. The INS structure and boundary conditions were determined on a national and global level from the point of view of resources, waste, infrastructure and human resources. At the second stage of the Joint Study assessment of the proposed INS CNFC−FR was performed using the INPRO methodology. A holistic approach of assessment was used for the reactor system as well as for all fuel cycle components from mining of raw materials to the final disposal of waste covering all areas of INPRO methodology, i.e. economics, safety, environmental protection, waste management, proliferation resistance, and infrastructure. In the course of the assessment, strong and weak points of the INS under review were determined. The outputs received during this stage shall contribute to further improvement and enhancement of the INPRO methodology. The results of the Joint Study received at the first stage of its implementation were summarized in a conference paper [6]. The preparation of the final report was implemented by members of the STC and experts of the INPRO international coordinating group (ICG).
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Distribution of responsibility and assistance in preparation of the first draft of the chapters of the final report are shown in Table 2.1.
TABLE 2.1. ACTIONS TO PREPARE THE FIRST DRAFT OF THE FINAL REPORT OF THE JOINT STUDY
N Preparation of a first draft Responsibility Assistance
1 Summary V. Usanov, ICG, IAEA 2 Chapter 1. Overview of the V. Usanov, ICG, IAEA V. Kagramanyan, STC, RF assessment results 3 Chapter 2. Outline of the study B. Raj, STC, India V. Usanov, ICG, IAEA 4 Chapter 3. National nuclear X. Mi, STC, China scenarios A. Vasile, STC, France B. Raj, STC, India R. Nakai, STC, Japan Y. I. Kim, STC, ROK S. Yugay, STC, RF 5 Chapter 4. Common and specific V. Usanov, ICG, IAEA STC Members features in national approaches 6 Chapter 5. Global context of the V. Usanov, ICG, IAEA STC Members assessment 7 Chapter 6. Economics S. Yugay, STC, RF Y. I. Kim, STC, ROK C. Loaec, France 8 Chapter 7. Resource A. Vasile, STC, France R. Nakai, STC, Japan sustainability and environment 9 Chapter 8. Safety of nuclear B. Raj, STC, India A. Vasile, STC, France installations 10 Chapter 9. Waste management A. Vasile, STC, France B. Raj, STC, India 11 Chapter 10. Proliferation Y. I. Kim, STC, ROK V. Kagramanyan, STC, RF resistance 12 Chapter 11. Infrastructure R. Nakai, STC, Japan A. Afanasyev, Ukraine 13 Chapter 12. Identification of B. Raj, STC, India. Mi, STC, China areas for collaborative R&D 14 Annex A. Outline of a French A. Vasile, STC, France J. Cazalet, F.L. Linet, Case Study D. Grenèche, France 15 Annex B. Japan’s evaluation R. Nakai, STC, Japan K. Ono, K. Sato, K. results Kurisaka, H. Shiotani, R. Aso, H. Kawasaki, A. Ohtaki, Japan STC = Scientific and Technical Committee of the Joint Study; ICG = International Coordination Group of INPRO; ROK = Republic of Korea; RF = Russian Federation
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The Joint Study has imbibed results of national INPRO assessment studies on an INS CNFC−FR performed by participants from several countries: Russian Federation (2004); France (2006); and Japan (2007). The STC of the Joint Study found it expedient to attach to the final report two Annexes: Annex A presents the generalized results of the French contribution to the Joint Study [7]; Annex B is to present to an interested reader, by the example of Japan, a full-scale national assessment on an INS CNFC−FR using the INPRO methodology.
2.4. DEVLEOPMENT OF THE INS CNFC-FR
The Joint Study considered a long-term perspective, i.e. 50−100 years. In this connection, a necessity was recognized to elaborate (to reach consensus) on some generalized characteristics of fast reactors suitable for all participants. Further it was agreed that these generalized characteristics should not be country-specific, but would also include nuclear energy systems with other types of fast reactors and thermal reactors. In developing a model of a CNFC−FR system, the following aspects were considered: • Operating, near term and innovative fast reactor designs of medium and higher installed capacity; • Modular reactors (sodium and lead-bismuth cooled fast reactors); • Divergence in the choice of fuels and coolants of fast reactors (sodium cooled reactors, lead cooled reactors, gas cooled reactors, oxide fuel, metallic fuel, carbide fuel, etc.); • Neutron balance of fast reactor systems as a physical basis for ensuring fundamental advantages of closed fuel cycle (enhanced uranium utilization, high efficiency usage of Pu with less risk of proliferation, burning of minor actinides, possibility of separating and storing long lived fission products safely); • Contemporary status of the nuclear fuel cycle in participating member states and plans aiming at its closure, namely reuse of reprocessing products (plutonium, uranium, minor actinides, etc.) in the nuclear fuel cycle and a decreasing amount of wastes and their radiotoxicity in the back-end of the nuclear fuel cycle; • Existing and planned future applications of fast neutron reactors (electricity, heat, water desalination, hydrogen production, etc.); • Nuclear fuel technologies: MOX fuel, nitride fuel, metallic fuel, pellet technology and vibro-pack technology, etc.; • Spent nuclear fuel reprocessing technologies (aqueous methods, non-aqueous methods); • Diverse approaches to the role of fast reactors: burners (consuming Pu) or breeders (producing Pu).
2.5. GLOBAL ENERGY SCENARIOS
The global scenarios considered in the Joint Study reflect the variation in the approaches with regard to nuclear growth and to the role of fast spectrum reactors vis-à-vis other reactor systems. Indeed, the uncertainty of nuclear energy growth in the forecasts of different energy agencies and organizations covers any conceivable possibility of nuclear power deployment – from significant growth to complete phase out. Similarly, there is no consensus regarding the future role of fast reactor systems. In global projections made in some countries, fast reactors contribution increases steadily beyond 2050 and fast reactors become the only energy source beyond 2150. The world nuclear capacity is predicted to stabilize at 5000 GW(e). In other projections of the global world made in other countries, the fast reactor contribution increases steadily beyond 2020 and advanced thermal reactors continue to play a significant role over a long period, and nuclear capacity increases monotonically. The Joint Study found a strong
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dependence of the judgment on the sustainability potential of a reviewed nuclear energy system on the selection of a global scenario − the higher the growth in the global nuclear scenario, the more the CNFC−FR could support a sustainable energy supply. The approach proposed by the study to reduce the level of scenario uncertainties was based on the compilation and analysis of national long term nuclear scenarios beyond 2050. It is evident that the efficiency of this approach depends on the size of the group of countries involved in the study and increases with its expansion. After a deliberate survey of a set of global scenarios of nuclear power development projected by different energy agencies and organizations, scenarios A2 and B2 of the special report on emission scenarios (SRES) of the Intergovernmental Panel on Climate Change (IPPC) [8] were selected by the participants of the Joint Study as consistent with their expectations of nuclear growth during the century. The scenario approach helped to evaluate the role of CNFC−FR systems in important areas of the assessment like assurance of nuclear fuel supply, waste management, and proliferation resistance. To consider issues related to the assurance of nuclear fuel supply, analysis of natural uranium consumption by thermal reactors with a once-through fuel cycle (OTFC) was performed for the global SRES B2 scenario using the computer code DESAE [9]. This code was available for all participants of the study. The analysis showed that in case of an increase of global nuclear capacities to 2000 GW(e) by 2050 the annual uranium extraction would reach a level of 300−350 thousand tons per year, while the cumulative uranium consumption by 2050 would exceed 10 million tons; this exceeds the sum of identified U resources (4.7 million tons) and prognosticated U resources (2.5 million tons) [10]. If the nuclear system’s capacity increases to 5000 GW(e) by 2100, annual uranium consumption would reach 750−900 thousand tons per year and integral uranium consumption would exceed 40 million tons; this is much higher than the 14.8 million tons of total conventional uranium resources documented in Ref. [10] or the assessments of U resources made by other competent organizations. Only the utilization of uranium from phosphate ore with production costs of 300 $/kg or more could meet this demand beyond the middle of the 21stcentury but the involvement of this resource may cause a significant rise of the uranium price and possible tensions in the world uranium market. Thus, the considerations taken in the Joint Study confirmed the concern of some countries related to fuel assurance and emphasized the need of timely introduction of plutonium and thorium based fuels. The scenario approach was applied in the Joint Study also to figure out the global radioactive waste inventory, industrial needs in enrichment, spent nuclear fuel reprocessing, MOX fuel fabrication, geological repositories and other parameters related to the multidimensional assessment of the INS.
2.6. NATIONAL NUCLEAR SCENARIOS
In the countries that participated in the Joint Study Natural, social and economic conditions differ to a great extent, and there is also little similarity in history, current status, and projected growth of nuclear power. The plans for introduction of fast reactors vary from country to country: by 2020 in India, 2020−2025 in Russia, 2030 in France, 2030−2035 in China and 2050 in Japan. In France and Japan the main objective introducing CNFC-FR is to enhance the sustainability features of their national nuclear power program assuming stable or slowly increasing nuclear capacities. China, India, Republic of Korea and Russia are going to use the full potential of the CNFC−FR system to enhance sustainability of their energy supply assuming high growth of nuclear power in their country. Fig. 2.4 illustrates the differences in nuclear growth projected in France and India.
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1975 2000 2025 2050 2075 Reactors Lifetime extension Gen IV Operating Fleet EPR a) France Source : EDF, ENC 2002 Cycle
U (Udep, URT) Recycling in Gen IV Gen IV Pu (recycling as MOX fuel) Cycle M.A. ---> glas or recycling M.A. + F.P. ---> Glas waste Storage F.P. ---> Glas
b) India
FIG. 2.4. Projected growth of nuclear capacities in France (a) and India (b).
The remarkable physic characteristics of fast reactors and the variety of options available for closure of their nuclear fuel cycle make it possible to adapt the INS CNFC−FR to specific national conditions and also to realize diverse aspirations of participating countries. For example, assurance of fuel supply through achieving high breeding factors in fast reactors is one of the main drivers for developing an INS in countries projecting fast growth of nuclear power from a current low level of installed nuclear capacities (China, and India). In contrast, a drastic reduction of high level radioactive waste requiring repository disposal by reducing the amount of uranium, plutonium, and minor actinides in waste to be disposed as well as more efficient utilization of uranium and plutonium are the main incentives for developing an INS in countries with a considerable nuclear share of the electricity sector and a low or moderate expected growth of nuclear capacities (France, and Japan). In case of Russia and Republic of Korea, where high growth of nuclear energy is planned to be continued from a rather significant level, both waste reduction and fuel assurance are important reasons for deployment of CNFC−FR. There was a consensus among participants of the Joint Study regarding benefits from the joint analysis of worldwide trends with a view to adapting the national strategies to future global realities.
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2.7. OBJECTIVES ACCOMPLISHED
2.7.1. Consolidation of outlooks
The Joint Study contributed to a consolidation of outlooks of its participants on the future of the CNFC−FR. The study revealed a set of commonalities regarding the introduction and further deployment of INS CNFC-FR. The important commonalities are: • Inevitability of CNFC−FR in some countries with large nuclear programs (~50 GW(e) or more by 2050); • High emphasis on economy, safety, non-proliferation and environment issues; • Acceptance limits in safety, environment, waste management, proliferation resistance, and infrastructure; • A steady growth of installed nuclear capacity based on improved thermal reactors; • A series of fast reactors to be deployed after a demonstration of a first of a kind fast reactor; • Sodium cooled fast reactors as a near term fast neutron system favored by current technologies; • R&D areas. Beside commonalities, participants of the study noted the following variants among national scenarios of the INS CNFC−FR development and deployment: • Priorities in timing of introduction of sodium cooled fast reactors; • Breeders and burners as variants of basically the same technology; • Choice of reactor types for development in the short term (pool or loop, capacity, breeding ratio); • Perspectives of long term reactor types and fuel cycle options (coolant, three or two circuits, advanced fuels); • Assessment of cost of recycling and alternate fuel cycles.
2.7.2. Identification of a reference CNFC−FR for the INPRO assessment
A specific task of the Joint Study was to develop a generalized model (technical specification) of the INS to be reviewed that has to imbibe the main features of the national systems of the participating countries without any duplication. After intensive discussions, a reference model of a near term INS CNFC−FR was identified (Section 4.5 of the report). Parameters of this reference model are based either on characteristics already demonstrated in operating nuclear installations or expected to be included in evolutionary designs. The chosen model is meeting one objective of the study that is to enable a screening assessment of the INS using the INPRO mehtodology.
2.7.3. Assessment of the INS CNFC−FR
A focal point of the Joint Study was a comprehensive screening assessment of the INS CNFC−FR for compliance with the requirements of sustainable nuclear energy supply as defined in the INPRO methodology. The assessment took into account results of assessments − using the INPRO methodology − of national fast reactor systems performed in the Russian Federation (2005), France (2006), and in Japan (2007). Based on an agreed methodological platform and boundary conditions, the Joint Study has considered the pros and cons of an
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introduction of fast reactors revealing new dimensions in understanding of the role of technological innovations for enhancing the sustainability of nuclear power. It was agreed to use as a reference model the near term INS CNFC-FR based on proven technologies – sodium coolant, MOX pellet fuel and aqueous reprocessing technology. Though representing only an initial phase of technology development, this system carries the main generic features of the technology. When assessing the INS CNFC−FR, the Joint Study focused not on emphasizing the differences between thermal and fast reactors, open and closed fuel cycles, but on finding a synergy among them that could improve the nuclear system as an integrated entity. In the course of the work, a context dependent character of the INS assessment was clearly recognized. The forecast on the growth of population, social and economic development, energy and nuclear demand, level of international cooperation, time horizon, and many other factors have a significant impact on the results of an assessment. The evaluation of the INS CNFC−FR was done in the Joint Study for a global scenario with a rather high nuclear growth. For this case, application of the INPRO methodology to review the INS indicated that all INPRO acceptance limits in all methodology areas except for economics were met. While the high evaluation of the INS sustainability features in safety, environment, and waste management − being in agreement with results of many other studies − does not seem surprising, the judgment of a high sustainability potential of CNFC−FR with regard to proliferation resistance and infrastructure aspects is an important finding of the Joint Study. As shown in the report, the advantageous qualities of the INS CNFC−FR in respect of proliferation resistance are determined by fundamental physical properties of a fast reactor. The chain ‘thermal reactor – fast reactor’ is capable of converting the spent fuel of a thermal reactor containing fissile material into valuable fuel for a fast reactor that eventually can be brought to an end-state unattractive for any kind of proliferation. In that way, fast reactors can balance accumulation and consumption of fissile material in a nuclear energy system by providing technical means for decreasing the proliferation risk associated with uncontrollable accumulation of fissile materials in spent fuel. Cost was found to be a weak point of the reviewed INS since the first of a kind fast reactor did not meet the INPRO economic requirements adequately. However, it was indicated by the participating countries that a large knowledge base exists. Particularly, accumulated knowledge reached both by “learning by searching” in the course of intensive R&D and “learning by doing” by construction and operation of demo and industrial installations, made it possible to decrease projected specific capital cost of sodium cooled fast reactors in recent projects close to the cost of water cooled thermal reactors (Fig. 2.5).
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Evolution of SFR/WTR spesific cost 3.5
3
SPX 2.5
2
1.5
Relative units Relative BN-600
EFR BN-800 PFBR 1 BN-1800 JSFR SFR/WTR=1 0.5
0 1965 1970 1975 1980 1985 1990 1995 2000 2005 Project year
FIG. 2.5. Evolution of ratio of specific capital cost of sodium cooled fast reactors (SFR) to water cooled reactors (WTR) in different countries. (BN = Russian fast reactors; EFR = European fast reactor; JSFR = Japanese sodium cooled fast reactor; PFBR = prototype fast breeder reactor in India; SPX=SUPERPHENIX in France).
The jointly performed assessment of economics clearly showed confidence that − based on the identified RD&D − a near term serial sodium cooled fast reactor using available technologies could be realized on a cost competitive basis in the countries mastering the fast reactor technology. In the whole, it was shown in a qualitative and quantitative way that a closed nuclear fuel cycle with fast reactors is an innovative nuclear energy system of high potential and credibility. It is further a realistic solution for achieving a sustainable nuclear energy system with limited available resources and for efficiently managing high level waste thereby reducing the requirement for a repository with a large size. At the same time, the Joint Study has concluded that only countries with sizeable nuclear power programme will have enough incentives to significantly increase the needed investment in RD&D and tolerate high investment risks. That’s why it is likely that the INS CNFC−FR introduction could prove to be more beneficial through creation of efficient multilateral collaboration between countries with and without CNFC−FR rather than through a rapid worldwide deployment of the INS CNFC−FR.
2.7.4. Identification of collaborative R&D work and priorities for possible cooperation
Prospects of collaborative RD&D projects have been defined in the following areas: • Capital cost reduction; • Economic challenges of advanced fuel cycles; • Use of probabilistic safety assessment (PSA) in the design of advanced fuel cycle facilities; • Seismic isolation technology in fast reactor design; • Technical basis and method to eliminate core disruptive accident (CDA); • Status report of carbide/nitride/oxide fuel for fast reactors; • Development of advanced clad and structural materials.
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Templates have been prepared throughout the performance of the Joint Study, which were applied to assess the different levels of technologies and commitment of different countries to introduce CNFC−FR on a commercial scale. The templates showed complementarity of national conditions that creates promising prerequisites for mutually beneficial long term collaboration. The Joint Study considered the issue of plutonium-based fuel as a part of the general problem of assurance of fuel supply. An example from the national study of China carried out with the use of the DESAE code shows that some scenarios of China’s nuclear power development lack the necessary amount of plutonium for fuelling the planned fast reactors (Fig. 2.6, left). At the same time, worldwide, a large amount of plutonium is expected to be accumulated in spent fuel of thermal reactors in case of realization of global scenarios like the SRES B2 scenario under the assumption of a small share of fast reactors in the world (Fig. 2.6, right).
FIG. 2.6. Plutonium balance in one scenario of China (left) and in the SRES B2 global scenario (right). The example of the plutonium balance considerations indicates a principal possibility to use multinational collaboration to mitigate both the problem of a deficiency of nuclear fuel supply in some countries of the world, and spent fuel accumulation in other countries. In general, the Joint Study participants have noted that no single country, taken separately, meets all the optimal boundary conditions for development of CNFC−FR systems (high energy demand, high level of nuclear technology and infrastructure maturity, high resource of nuclear fuel, etc.), while the global nuclear energy system does. Thus, it was concluded that untapped reserves for realization of the INS CNFC−FR potential still exist by international cooperation. In the framework of a proper legislative basis they could become a powerful leverage to MS’ nuclear capacity building in a sustainable manner. A list of possible areas of international cooperation on CNFC−FR was determined by participants of the Joint Study in accordance with a scheme shown in Fig. 2.7.
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IAEA: INPRO MS: INS Methodology Development
INS Assessment
Identification of areas for INS improvement & cooperation, establishment of frameworks
Collaborative Projects
FIG. 2.7. A typical realization of the objectives of the INPRO Joint Study.
A few issues from the list were selected as proposals for INPRO collaborative projects to be initiated at the INPRO steering committee by certain Member States and implemented in the second phase of the INPRO project (Table 2.2).
TABLE 2.2. COLLABORATIVE PROJECT PROPOSALS N Collaborative Project Initiating country 1 Modeling, benchmarking and validation of safety grade decay heat India removal system for liquid metal fast reactors 2 Global architecture for INS based on thermal and fast reactors Russian including closed fuel cycles (GAINS) Federation 3 Proliferation resistance analysis and methodology improvements for France multinational INS (PROMM) 4 Assessment using the INPRO methodology on the basis of general Russian characteristics and input/output parameters of advanced and Federation innovative NFC within a large scale INS based on CNFC concept to satisfy principles of sustainability in the 21st century (FINITE) Subsequently, all collaborated project proposals produced by the Joint Study were endorsed by INPRO steering committee for implementation (GAINS and PROMM were merged into an enhanced GAINS project).
2.8. OBJECTIVES PARTLY ACCOMPLISHED AND NOT ACCOMPLISHED
The INPRO methodology contains more than a hundred indicators to be evaluated in order to come to an overall judgment on an INS’ compatibility with requirements of sustainable energy supply. Quantification of all indicators for the reference model of a near term INS CNFC−FR has not been completed.
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More emphasis on the fuel cycle is necessary, since innovations in the fuel cycle could significantly contribute to the growth of fast reactors implementation. A reference model of a future INS CNFC−FR with enhanced sustainability features that is expected to replace the near term CNFC−FR could not be developed since there is no agreement presently on the main technical decisions for such a system. The need for such a model will evidently arise in case of addressing a dynamic simulation of transition scenarios up to the end of 21st century. The Joint Study has shown that interactions among various groups interested in R&D should be intensified. There is a potential for improvement of the Joint Study results by way of enhanced participation from additional Member States in follow-up studies. Some of the objectives yet to be accomplished are: identification of infrastructure and readiness to share the infrastructure, assessment of fuel cycle facilities in various aspects, in particular economy and safety, and assurance of finance for future R&D activities.
2.9. FEEDBACK TO THE METHODOLOGY IMPROVEMENT
Detailed consideration of advantages and drawbacks of the INPRO methodology is given in the chapters of the report related to specific areas of the INS assessment. In this section the comments on the methodology are briefly summarized.
2.9.1. INPRO methodology as a step in developing a comprehensive instrument for INS assessment
The INPRO methodology for the assessment of innovative nuclear reactors and fuel cycles is considered by the participants of the Joint Study as a significant advance in development of an instrument for comprehensive analysis of ways enhancing sustainability features of nuclear power. The procedure for an INS screening assessment is well formulated and documented. Several propositions of the methodology were found to be especially helpful as a guide in making assessment of the INS CNFC−FR: an application of an interdisciplinary approach; consideration of the whole life time (cradle to grave) of the system; gradation of maturity levels; evaluation of not only a competitive potential of the INS but a potential of synergy with thermal reactors as well. Participants of the Joint Study are convinced that the advantages of the INPRO methodology identified in this study are far from covering its true worth. Assessors of other nuclear systems could add more items to the list of the merits of the INPRO methodology.
2.9.2. Identified ways for the methodology improvement
Along with merits of the INPRO methodology, this study identified several directions for its further improvement. Among them are: • Standardisation of acceptance limits (AL) via international harmonization of their values; • Introduction of additional ALs for fuel cycle facilities; • Development of a more definite guidance in providing adequate conditions of comparison of an innovative nuclear energy system (INS) with an existing nuclear energy system (NS); • Enhancing of the discriminative power of the assessment method through introducing of a target value scale in which ALs should be considered as the nearest target value; • Elaboration of an approach for consolidation (aggregation) of judgments obtained for specific INPRO variables (indicators);
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• Transition from assessing the capability of an INS to be in compliance with sustainability criteria (just to have entered the sustainability area) to enabling the segregation of alternatives within sustainability options and the selection of a nationally preferable design and a strategy of CNFC−FR development. The first two points of the list above seem to be rather clear while the following four points worth brief comments. A more definite guidance is needed in providing of adequate conditions for comparison of innovative nuclear systems with existing ones and should be included in an advanced assessment method to assure comprehensive assessment of nuclear systems with different levels of maturity. Citing and analysing of the scale of R&D investments needed to master different components of an innovative nuclear technology, of typical overdesign problems and the growth of initial projected costs, of delays connected with licensing and construction, etc. could be relevant and a useful part of a future manual how to apply the INPRO methodology for an assessment of an INS. The joint study has emphasized an expediency to increase the discriminative power of the INPRO methodology with an aim to provide a better navigation among the ways of nuclear system development and to prioritize R&D to be carried out. The Joint Study showed, using the INPRO methodology as currently developed, it was not possible to discriminate scenarios implementing very different levels of high level waste (HLW) transmutation and assess their environmental benefits in the back end of the INS CNFC−FR. Based on analysis, high level waste transmutation would allow reducing the radio-toxicity inventory and the heat generation in the final repository, and could have a non-negligible impact on characteristics of waste. However, based on the INPRO criteria documented in the current INPRO manual, better characteristics of waste achieved by HLW transmutation does not improve the judgment (using the INPRO methodology) regarding health and environmental qualities of the assessed system. This is caused by the fact that all positive changes of characteristics that happen due to the application of transmutation strategies lay in the tolerance range where characteristic of waste are better than the INPRO acceptance limit. Introduction of target values for INS indicators which could characterize the scale of improvements is a well-known approach used in some methods of the management science. An expediency to define target values was also shown in the report of Phase 1B of INPRO [2] but it has not been developed in the current manual into quantitative values. Results of the Joint Study on the INS CNFC−FR have confirmed in many assessment areas an expediency to increase the discriminative power of the assessment method and at the same time have denoted a need for aggregation of judgments made on different INPRO indicators. When improving of some characteristics of an assessed system leads to degradation of other ones, the necessity becomes rather evident to enable a balance of assets and liabilities in the system development strategy by introducing of a quantitative measure of significance (“weight”) for each INPRO indicator. The Joint Study has demonstrated a country specific focus in the development of an INS CNFC-FR that is being achieved through enhancing specific parameters of the INS. Introduction of significance (weights) for INPRO indicators based on national priorities would help to better understand different approaches of design and technological strategies of the CNFC−FR development. Eventually, introduction of “weights” for indicators of sustainable energy supply would give an opportunity to identify a justified balance between nuclear and non-nuclear energy options in the future national and global energy mix.
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The overall judgment of the Joint Study on the INPRO methodology adds up to a conclusion that the methodological platform built during the first phase of INPRO is a necessary but not a sufficient condition to solve the challenging task of producing a sound method for a comprehensive assessment of an INS. Hard and permanent work will be still needed to reach a higher discriminative power of the INPRO methodology. Since the worth (cost) of decisions in R&D in nuclear industry is very high, there is no doubt that the work on further improvement of the INPRO methodology and related tools will be continued at a national level. It seems reasonable that the IAEA would support these activities on a regular basis with the aim providing to the Member States an internationally agreed methodology and a toolset helping to harmonize national strategies to reach enhanced safety, environmental friendliness, proliferation resistance and economic viability of global nuclear power.
2.10. ADDED VALUE
The Joint Study has helped to establish an innovative and unique multinational organizational structure to develop INS based on CNFC−FR. The organizational structure has demonstrated its effectiveness and fitness for purpose. The multinational input available now would help to fine tune the INPRO methodology to identify a large INS CNFC−FR with high relevance to the world with regard to sustainability of energy supply. The Joint Study is a rational approach to define innovative R&D of mutual interest among various participants. The approach needs to be enhanced to identify the commonalities and variants of CNFC−FR with the goal to choose the best solution for the future. Robust mechanisms should also be spelt out for close interactions with stakeholder groups. The issues raised by the Joint Study addressed the broad interests of the IAEA Member States. The study has become a launcher for a few IAEA/INPRO collaborative projects related to different INPRO activities: modeling, benchmarking and validation of a fast reactor safety system; assessment of advanced and an innovative nuclear fuel cycle as a subsystem of the INS CNFC−FR; a study of the architecture of a global INS based on a CNFC with thermal and fast reactors capable to meet the requirements of sustainable energy supply and also to provide an overview of legal, economic, and technological incentives for cooperation.
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CHAPTER 3 NATIONAL NUCLEAR SCENARIOS
In this chapter the countries participating in the study outline their national nuclear scenario.
3.1. CHINA 3.1.1. Brief history and current status of the fast reactor technology development Basic research (1968−1987) The basic research for fast reactor technology were started in the mid 1960ies, the emphasis was put on fast reactor neutronics, thermo hydraulics, sodium technology, fuels, materials, sodium instrumentations and equipment in a small scale. Up to the year 1986, about 12 sodium loops and facilities had been built and used for experiments, including a fast neutron zero power facility, a sodium corrosion test loop, a sodium purification loop, a sodium thermo hydraulic loop, a sodium convection corrosion loop, a control rod driven mechanism parts testing facility, a sodium plugging meter testing loop, a stress corrosion sodium loop, an alternating magnetic pump sodium testing loop, a direct magnetic pump sodium testing loop, a sodium pump testing loop, etc. Applied basic research (1988−1993) In the framework of the national high-tech program applied basic research had been conducted with the construction of China experimental fast reactor (CEFR) as the target. The main emphasis was put on a fast reactor design study, sodium technology, materials, fuels, safety, etc. Main loops and facilities include a sodium purification loop, a sodium boiling testing loop, a mass transfer sodium loop, a fission products-cladding testing facility, a creep- fatigue sodium testing loop, a bi-axial creep testing facility, a U-Zr induction heating molten and pressure casting testing facility, an alloy fuel molten and pressure casting facility, a Na- H2O reaction testing loop, a hydrogen detection sodium loop, a carbon analysis facility, etc. China experimental fast reactor (CEFR) In the framework of the national high-tech program the China experimental fast reactor (CEFR) project has been implemented since 1990. The CEFR is a sodium cooled 25 MW(e)/65 MW(th) experimental fast reactor with (Pu, U)O2 as fuel, but UO2 as the first loading, Cr-Ni austenitic stainless steel as fuel cladding and reactor block structure material, bottom supported pool type, two main pumps and two loops for primary and secondary circuits, respectively. The water-steam tertiary circuit is also composed of two loops but the superheat steam is incorporated into one pipe connected with a turbine. The general timetable is shown in Table 3.1. TABLE 3.1. TIME TABLE FOR DEVELOPMENT OF CHINA’S EXPERIMENTAL FAST REACTOR (CEFR)
Conceptual design 1990−1992.7 Consultation with Russian FBR association and optimization 1993 Technical Co-design with R-FBR-A 1994−1995 Preliminary design 1996−1997 Detail design 1998−2003 Preliminary safety analysis report review 1998.5-2000.5 Architecture construction (first pot of concrete) started 2000.5 Reactor building construction completion 2002.8
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The reactor core is composed of 81 fuel subassemblies. There are three safety subassemblies, three compensation subassemblies and two regulation subassemblies; further there are 336 stainless steel reflector subassemblies and 230 shielding subassemblies and in addition 56 positions for primary storage of spent fuel subassemblies. The CEFR block is composed of a main vessel and a guard vessel supported from the bottom on the floor of the reactor pit with a diameter of 10 m and a height of 12 m. The reactor core and its support structure are supported by lower internal structures. Two main pumps and four intermediate heat exchangers are supported by upper internal structures. These two structures sit on the main vessel. Two heat exchangers of the decay heat removal system (DHRS) are hung from the shoulder of the main vessel. The double rotation plugs on which control rod driving mechanisms, a fuel handling machine and some instrumentation structures are mounted are located on the neck of the main vessel. After the preliminary design, 40 items had been selected to be experimentally verified including the residual heat removal system, a siphon-break facility, and structure erosion testing, etc. Along with the design, some design demonstration loops and facilities have been built. In order to supply nuclear-grade sodium of the CEFR, a sodium purification technology has been developed. The first step was to construct a reduced scale purification facility with a capacity of 300 kg/day. The quality (purity) of sodium achieved by this facility is less than about 28 ppm impurities. Based on experience of the reduced scale facility, a large scale facility has been constructed near Wuhai city with a capacity of 1500 kg/day. The sodium purity index is below 20 ppm impurities. The CEFR is under construction right now. Most of the systems are being installed. The main vessel is close to be finished. The pressure testing of the vessel has been completed. According to the installation schedule, all of the systems below +16.8 m of the main building should be completed before the end of 2006. The conventional island and most of the auxiliary systems, such as ventilation systems and water systems should be finished in 2006. The commissioning of some auxiliaries systems has been finished, such as the deionized water workshop, gas station, vacuum systems, sodium-acceptance system, etc. And, up to now 250 tons of sodium have been transported to the site and stored in sodium tanks. According to the reactor project plan, the reactor should reach first criticality in the middle of 2009 and connect to the net in the middle of 2010. China prototype fast reactor (CPFR) A core design study has been started in 2005 for the CPFR which is the second step of fast reactor engineering development in China. The power of the CPFR will be 600 MW(e) and it is intended to keep its main technical features the same as the CEFR as much as possible.
3.1.2. Incentives and boundary conditions for the future deployment of INS CNFC-FR
Since the end of 1970ies, China’s national economy has been entering into a stage of continuously stable development. During the past five years, the average rate of increase of the national gross domestic product (GDP) was 9.4 % as shown in Table 3.2. Development of the economy was supported by an energy supply with a corresponding growth. The annual primary energy consumption for the period of 2001−2005 is also shown in Table 3.2.
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TABLE 3.2. NATIONAL ECONOMY DEVELOPMENT IN CHINA
Year 2001 2002 2003 2004 2005 GDP Annual Increasing Rate (%) 8.0 9.1 10.0 10.1 9.9 Annual Primary Energy Consumption 1.43 1.52 1.75 2.03 2.23 (Billion tce) The total primary energy consumption in 2005 was 2.23 billion tce which is composed of different components as shown in Table 3.3.
TABLE 3.3. PRIMARY ENERGY CONSUMPTION IN 2005
Energy Type Consumption Growth rate from 2004 Coal (t billion) 2.14 10.6 % Oil (t billion) 0.3 2.1 % Gas (m3 billion) 50 20.6 % Hydro (TWh) 401 13.4 % Nuclear (TWh) 52.3 3.7 % According to the goal for national development, China’s GDP should double from 2000 to 2020 resulting in a corresponding growth of energy demand. To satisfy the demand for electricity in 2020 the installed capacity of power plants should increase from 400 GW(e) to a range of 960 to 1000 GW(e). More than half of this increase is to be filled by coal fired plants and about a third is to be contributed by hydro plants. The remaining power necessary should be supplied primarily by nuclear power, i.e. the nuclear capacity should be increased from the currently 9 GW(e) to about 40 GW(e) in 2020. In 2009, China probably will adjust the goal of nuclear power development to 70 GW(e) in 2020. China’s energy security is facing to two major challenges. The first one is the contradiction between an ever increasing demand for energy and the insufficient resources of fossil fuels. Presently, China’s per capita energy consumption is only about 1 tce/a, which is about half of the world average. China’s energy supply has to be increased significantly in the coming decades to meet the requirements of the rapid development of its national economy. The second challenge comes from an irrational energy supply structure, in which coal covers 60 % of the primary energy and 75 % of the power generation. The coal based energy supply structure causes serious problems of environmental pollution and emission of greenhouse gases (GHG). China has become one of the most polluted countries in the world. In the year 2000 as an example, China emitted 20 Mt of SO2 and 3 Gt of CO2, both ranked as world No.2. The economic loss by environmental pollutions accounts for 3~7 % of GDP in China. These problems could hinder the further development of the country if they would not be solved properly. With the drastic increase of China’s oil and gas consumptions since the mid 1990ies, China is becoming more and more dependent on import of oil and gas, and the major source of oil imports is the Middle East. This has forced the people to pay attention to the topic of energy security in China. Owing to the rapidly increasing demand of oil and gas, the fact that China’s economy is heavily dependent on oil and gas imports seems difficult to be reversed in the future. Data show that in 2004 China’s oil import was over 100 Mt which covered about 30 % of the total oil consumption. It is estimated that by 2020, China’s oil demand will be 450 Mt and more than 50 % of supply will depend on import. Facing this severe situation, some people suggest to practice a multi-channel import and to establish national strategic reserves of oil. These
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measures may be effective to alleviate some problems in the next decades. But, considering the fact that the oil and gas reserves in the world could be depleted in less than 100 years, from now on the energy structure in China has to be adjusted, gradually reducing the share of coal, oil and gas and increasing the share of new energy resources. Among alternative energy resources, hydro power resources, as renewable and clean energy resources, are abundant in China and will be exploited with top priority in the coming years. In China, the economically exploitable hydro-power resource is estimated to be about 400 GW(e). At present, the installed capacity of hydro-power is some 100 GW(e). The non-hydro renewable resources, such as wind, solar, and biomass, are highly encouraged to be developed (especially wind power). However, they could not replace the fossil fuels on a large scale in the foreseeable future owing to their low energy density, intermittent supply and high cost. Considering the fact that the existing energy structure in China is based on fossil fuels (especially coal), China’s near-term energy policy (before 2020) has to be dependent on coal as the major energy source while actively developing hydro power, nuclear power and other new energy resources and putting top priority on improving energy efficiency. This calls for an accelerated development of nuclear power as fast as possible. As a well-developed clean energy with zero discharge of GHG, nuclear energy has been received more and more attention in China and is regarded as a big supplement to fossil fuels together with hydro power. There is a big room for the development of nuclear energy in China. Nuclear Energy will be one of the major energy sources in the coming 50 years in China. With the rapid growth of national economy, it is expected that by the year 2020 China’s total energy demand would be increased from the present 1.3 Gtce to 3.0 Gtce or more and the power capacity would be increased from the present 400 GW(e) to some 960～1000 GW(e) (at least a 560 GW(e) increase). More than half of the 560 GW(e) increase will be contributed by coal-fired power, and the installed capacity of hydro-power would be increased by more than 170 GW(e). Even then, there is still a big gap to be filled. This calls for the accelerated development of nuclear power as fast as possible. China’s national goal by 2020 is that nuclear power capacity will be increased from the present 9 GW(e) to 40 GW(e) in operation with another 18 GW(e) under construction. The nuclear share will be some 4 % of the nation’s total power capacity by 2020. After 2020, it is predicted that China’s domestic supply of oil and gas will drop and will become mainly dependent on import; a larger amount of use of coal will be limited by the environmental burden (the highest scale of coal burning should be less than 3.4 Gtce/a); there will be no big room for further expansion of hydro-power after its capacity reaches 300 GW(e) owing to the limited economically exploitable hydro resources in China (< 400 GW(e)). Other renewable resources (such as wind and solar energy) will be fully encouraged to be developed and will play an increasing role, especially in some remote or coastal areas. But it will be unlikely that the renewable energy resources could replace the fossil resources on a large scale in the foreseeable future. A major way to meet energy demand after 2020 is thus to further increase the share of nuclear energy so as to ensure both energy security and environmental safety of the country. By the mid of 21st century, the capacity of nuclear power in China should reach ~240 GW(e) or even higher with its share increased to 15 % of the total power capacity. Nuclear power will then be one of the three major energy resources in China together with “cleaned” coal and hydro-power; thus, nuclear energy will make greater contributions to China’s sustainable development.
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According to the world trend of nuclear power development and an ever increasing energy demand in China, it is necessary to develop advanced nuclear energy systems in China. China will develop and deploy nuclear power plants with advanced pressurized water reactors in the coming 30 years or so and also fast reactor power plants, together with the development of the associated nuclear fuel cycle technologies, so as to meet the requirements of sustainable development of the nation’s economy and environment. For a sustainable nuclear energy supply in China, stepwise installations of thermal reactors−fast reactors−fusion reactors have been decided as a basic way. For nuclear fission energy, fast breeder reactors will play an important role using uranium resources efficiently. Fast burner reactors will transmute minor actinides, and more efficient burners based on accelerator driven systems will minimize volume and toxicity of high level waste which needs geologic disposal.
3.1.3. National planning of capacities and structure of energy sector
According to the study in the energy area of the national high-tech program more than ten years ago, the envisaged annual primary energy production in China for 2050 totally would reach 4.5 billion tce as shown in Table 3.4. And a recent study shows that it will be 5.0 billion tce. Nuclear capacity should be developed up to 240～250 GW(e) by 2050.
TABLE 3.4. ENVISAGED PRIMARY ENERGY SUPPLY IN CHINA FOR 2050
Energy supply envisaged in Energy supply envisaged in 1991 2005 Energy Exploitable Standard Total Standard Total source In 2050 coal requirement Coal requirement equivalent (billion tce) Equivalent (billion tce) (billion tce) (billion tce) Oil 0.1×109 t 0.5 Gas 1500×109 m3 0.45 0.3 Hydraulic 260~370 GW(e) 0.65 0.6 Coal 3.4×109 t 2.50 2.5 Nuclear 240 GW(e) 0.60 0.6 Others 0.30 0.5 Total 4.5 4.5 5.0 5.0 Based on Table 3.4 China needs about 250 GW(e) nuclear capacities by 2050. For such a huge capacity it is impractical to use only pressurized water reactors (PWR) to meet it, due to PWR needs of about 2.4 to 2.5 million tons uranium for a 60 years operation. Although China is abundant in uranium resources theoretically, explored uranium reserves of technically feasible and economically affordable are limited. And, only about 4~5 million tons reliable uranium resources are estimated in the world with a price lower than $ 130/kgU, which is equivalent to 2.6 times of the present price. Regarding the nuclear share in China’s mainland, nuclear power development was initiated in the early 1980ies. After the first nuclear power plant (NPP) Qinshan-I with a 300 MW(e) PWR based on domestic technology started operation in 1991, the country has started to use the nuclear energy in locations with lack of fossil resources as a subsidiary. Up to now nine NPP units have been showing good operation records.
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For nuclear power development, the following two very important remarks have been made in the sketch of national mid-long term science and technology development program (2006−2020) issued by the State council of the People’s Republic of China on February 9th, 2006: (1) “…to energetically develop the nuclear power technology and to form a R&D capability of our own on nuclear power technology…” (2) “…the advanced nuclear energy system Generation IV, nuclear fuel cycle and fusion energy technology are paid more and more attention …” Under this sketch of a program the government has decided to develop continuously nuclear power with a mid-term target of 40 GW(e) in 2020. The Government has approved 4 new NPPs with 8 PWR Units. Additionally 30 NPP units at 7 sites are planned. Sustainable development of nuclear power depends on the full utilization of uranium resources and the minimization and proper disposal of nuclear waste. Considering the fact that the economically exploitable uranium reserves in the earth crust are limited (~4.6 Mt with costs <130 $/kg), sustainable development of nuclear fission energy depends on the fast breeder reactor (FBR) energy system, which is based on a closed fuel cycle. China’s nuclear energy development will gradually introduce a thermal reactor system–fast reactor system. It is estimated that at least 30~40 years are needed for commercialization of a FBR energy system in China. Therefore, China will have to deploy almost exclusively thermal reactor power plants (mainly PWRs) before 2030 and PWRs will probably be continued to be installed after 2030 as long as cheap uranium is available.
3.1.4. The optimistic and moderate scenarios for development of the CNFC with fast reactors
The nuclear power program in China foresees a significant increase of nuclear power by 2050 and a further increase thereafter until the end of the century. The power is to be primarily generated with PWR until about 2030~2040 reaching a value of 100~150 MW(e) (a value kept constant thereafter). Beginning in 2020, fast breeder reactors are planned to be installed and become the primary source of nuclear electricity by 2050 generating about 100~130 GW(e). Two nuclear power scenarios (Fig. 3.1 and Fig. 3.2) have been selected for the assessment, supposing a certain amount of natural uranium resource would be available. In China a complete nuclear fuel cycle with all front end and back end facilities are part of the national nuclear energy system.
In the actual assessment using the INPRO methodology – as done by all participating countries in the Joint Study – a fleet of fast breeder reactors together with a closed fuel cycle was evaluated.
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FIG. 3.1. Simulated nuclear energy development scenario in China (moderate case).
FIG. 3.2. Simulated nuclear energy development scenario in China (optimistic case).
3.1.5. National approach to identifying fast reactors designs and relevant fuel cycles
Plan of development of fast reactor engineering and its fuel cycle According to a strategic study on the fast reactor development in China, the fast reactor engineering development will be divided into three steps: 1. China experimental fast reactor (CEFR) with a power 20 MW(e)/65 MW(th); 2. China prototype fast reactor (CPFR) with a power 600 MW(e)/1500 MW(th); 3. China demonstration fast reactor (CDFR) with a power 1000~1500 MW(e)/ 2500~3750 MW(th).
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The CEFR has been implemented under the framework of the national high-tech program. The reactor is now under construction and pre-operation testing and its first criticality will be achieved in 2009. The CPFR is proposed by the government as an important project of the eleven’s five year plan (2006−2010) with the target to complete its construction in 2020. In order to shorten the development period from CEFR to CDFR and to decrease technical- economic risks, maximum continuity will be emphasized in the main technical selections for these two reactors, such as coolant, primary circuit structure, decay heat removal system, fuel handling systems, etc., as shown in Table 3.5. The demonstration fast reactor will be constructed around 2035. TABLE 3.5. TECHNICAL CONTINUITY IN THE DESIGN OF CHINESE FAST BREEDER REACTORS CEFR CPFR CDFR Power, MW(e) 25 600 1000～1500 Coolant Na Na Na Type Pool Pool Pool UO MOX MOX Fuel 2 MOX Metal Metal Cladding Cr-Ni Cr-Ni, ODS Cr-Ni, ODS Core Outlet Temp, t°C 530 500～550 500 Linear Power, W/cm 430 450～480 450 Burn-up, MWd/kg 60～100 100～120 120～150 DRPs DRPs DRPs Fuel Handling SMHM SMHM SMHM IVPS IVPS IVPS Spent Fuel Storage WPSS WPSS WPSS ASDS ASDS+PSDS ASDS+PSDS Safety PDHRS PDHRS PDHRS
ASDS: Active Shut-Down System; DRPs: Double Rotating Plugs; IVPS: In-Vessel Preliminary Storage; ODS: oxide dispersion strengthened steel; PSDS: Passive Shut-Down System; PDHRS: Passive Decay Heat Removal System; SMHM: Straight Moving Handling Machine; WPSS: Water Pool Secondary Storage. For the fuel cycle, MOX fuel will play an important role in the start and intermediate stage of fast reactor development. In the optimistic scenario of fast reactor development, the usage of metal fuel is necessary. Metal fuel meets non-proliferation requests; additional proliferation resistance is achieved in an integral fast reactor with an on-site pyroprocessing fuel cycle. The Chinese fuel cycle system is developed step by step. China’s spent fuel reprocessing pilot plant with a capacity of 50~100 t(HM)/a is under construction and will be put into trial operation by 2009. A commercial reprocessing plant is under consideration and expects to be built by around 2020. The development of MOX fuel fabrication in China is now in an early stage. A 500 kg/a MOX fuel fabrication experimental line is under construction and will be put into operation by 2010. This line will produce test fuel for the CEFR firstly and driver fuel for CEFR later. The test fuel for the CPFR will also be produced by this experimental line. A
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MOX fuel fabrication plant with a capacity of 40～100 t/a is under consideration with the aim of commissioning by 2020. Near-term main activities for fast breeder reactors with CNFC 1) China experimental fast reactor CEFR: • Time period 2006−2010 • Continued installation up to 2008; • Continued pre-operation testing up to 2009; • Collection of data during pre-operation testing and trial-operation and verification of design computer codes and to safety analysis codes, connected with neutronics, thermo hydraulics, shielding, mechanical system analysis and transient analysis; • Identification of safety regulations, design criteria, safety guidelines and standards used for CEFR, which were transferred mainly from light water cooled reactors; • Safety experiments at the CEFR, designed with care, UTOP, ULOF, ULOHS, performance of DHRS; • R&D activities to support CEFR safety operation, as follows: o CEFR simulator for operator team training; o Equipment maintenance simulators; o Fast reactor design data files; o Sodium CEFR-sized test loops design and construction for product qualification of some key components; o Advanced liquid metal laboratory; o Long term irradiation performance testing of materials used by CEFR; o Irradiation capsule technology for materials and/or different fuels irradiated in CEFR; o Quick-welding technique for sodium leakage from piping; o Caesium trap study; o Spent sodium treatment technology; o Establishment of some expert systems(operator support system, transient state operation optimization system, core fuel management optimization system); o All-digital surveillance and control system development. • Time period 2011−2020 • Continuous safety experiments; • Irradiation experiments and PIE study: o Materials: 316Ti, 15-15Ti, HT9, ODS; o Fuels MOX, U-Pu-Zr, (Pu, MA, UO2). • Accumulation of CEFR operation and maintenance; • Fabrication localized key components used in CEFR; • Operation data accumulation; • R&D activities to support CEFR safety operation, as follows: o Reformation and supplement of PIE hot cells; o Irradiation capsules design fabrication and demonstration; o Diverse analysis support; o Advanced operation support system development; o Development of on line inspection techniques.
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2) China prototype fast reactor CPFR • Time period 2006−2011 • Design study: o Support style of main vessel: top hung or bottom supported; o Loop number of secondary circuit; o Reactor building seismic isolation (or not). • Conceptual design; • Preliminary design of nuclear island; • Site selection study; • Feasibility study for project; • R&D activities to support CPFR, as follows: o Design and construction of a zero power facility; o Design and construction of a multi-purposed sodium engineering loop mainly for sodium pump and steam generator with 600 MW(e); o Design and construction of a water mock-up for the CPFR reactor block for a DHRS study; o Advanced design computer codes and safety analysis codes development. • CPFR key components design and trial-fabrication study: o Sodium pumps; o Steam generator; o Control rod driven mechanism; o Fuel handling machine; o Passive shut-down system; o Delayed neutron detection system; o Advanced purification system for primary sodium. • Time period 2011−2020 • Preliminary design and detailed design; • CPFR construction, pre-operation testing up to commissioning by 2020; • R&D Activities to support CPFR, as follows: o Completion of design demonstration: key components, passive shut-down system, passive decay removal system, anti-seismic isolation device demonstration (if to be used); o Completion of zero power facility ; o Concept study to use Pb-Bi for secondary circuit coolant; o Completion of design demonstration proposed from simplified primary sodium purification system. 3) R&D activities to support CNFC • Time period 2006−2010 • Construction of a PWR spent fuel reprocessing pilot plant with a capacity of 50~100 t/a; • Construction of a CEFR MOX fabrication laboratory with a capacity of 500 kg/a; • MA extraction process technology development, combined with PUREX process; • Advanced materials for fast reactor core structure and steam generators, trial- fabrication and testing: 15-15Ti, ODS, HT9, 9Cr1Mo. • Pyro-process and injection casting technology study; • 1.2 t/a pyro-process and injection casting facility design.
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• Time period 2011−2020 • R&D activities to support CNFC, as follows: o (Pu-MA-U)O2 fuel fabrication; o Key equipment fabrication with a 1.2 t/a pyro-process and an injection casting complex; o Design demonstration for a 800～1000 t/a or 200 t/a reprocessing plant; o Design demonstration for a 40~100 t/a or 10~15 t/a MOX plant. Current international cooperation on fast reactors with CNFC • IAEA-INPRO-Joint Study: o Fast reactor design study: reactor building seismic isolation, reactor vessel support selection (hanging or sitting), two or three loops for secondary circuit, ‘Curie point’ passive shut down system, immersed primary sodium purification in reactor vessel; o CNFC: To exchange the MOX and (U,Pu-MA)O2 fuel design experience, to learn from other member states the fabrication experience, of (U,Pu-MA)O2 and MOX fuel with high Pu percentage (～40 %), to share the experience of irradiation tests; o Pyro-chemical process of spent metallic fuel: the most promising options for reprocessing of spent U-Pu-Zr fuel, the design and construction of lab-scale set- ups both for cooled tests and hot cell tests, the selection of a most suitable molten salt system, the expertise of efficient heating systems, the use of corrosion- resistant materials, the fundamental data on pyro-electrochemistry including the basic thermodynamic and kinetic parameters. • Russian Federation: o Consultancy on CEFR construction and pre-operation testing. • France o Fast reactor R&D cooperation under the framework of agreement of research reactor R&D cooperation between CAEA and CEA. • Japan o Fast reactor safety analysis and operator training under the framework of MEXT program between China and Japan.
3.2. FRANCE
3.2.1. National strategies and scenarios for development of a closed fuel cycle with fast reactors (CNFC−FR)
Present French nuclear system
France has 59 nuclear reactors operated by Electricité de France (EdF) (58 PWRs and the PHENIX fast reactor), with a total capacity of over 63 GW(e), supplying over 426 billion kWh of electricity per year, 78 % of the total electricity generated in France. In 2005 French electricity generation was 549 billion kWh net and consumption 482 billion kWh–7700 kWh per person. Over the last decade France has exported 60–70 billion kWh net each year. The French PWRs are of three types: 900 MW(e), 1300 MW(e) and 1450 MW(e) (N4 type) and were launched during a relatively small period of time as shown in Fig. 3.3.
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MW 70000 60000 N4
50000 40000 1300 MW 30000
20000 900 MW 10000 0 1975 1980 1990 2000 2005
FIG. 3.3. Schedule of French PWR deployment. Early in 2003 France's first national energy debate was announced. The debate was to prepare the way to define the energy mix for the next 30 years in the context of sustainable development at a European and global level. The role of nuclear power was central to this debate, along with specific decisions concerning the European pressurised water reactor (EPR), and defining the role of renewable energies in the production of electricity, and in thermal use and transport. In May 2006 the EdF board approved construction of a new 1630 MW(e) EPR unit at Flamanville, Normandy, alongside two existing 1300 MW(e) units. The decision is seen as "an essential step in renewing EDF's nuclear generation mix".
Fast reactors development [11]
In the 60ies, France started a programme of development of fast reactors (Fig. 3.4). The main incentive was to master a technology able to use all the energetic potential of uranium, and possibly of thorium; the oil crisis in the 70ies in conjunction with perspectives of growing energy needs made it desirable to rapidly reach the industrial stage of this development programme. From RAPSODIE to SUPERPHENIX and EFR, this development is remarkable by its coherence and its continuity. It finally led to the demonstration of the viability of oxide fuelled-sodium cooled large fast reactors, and the essential knowledge to design optimised plants. Most of this work was accomplished in the frame of an European collaboration. RAPSODIE The first studies of the RAPSODIE reactor started in 1958 and construction of the reactor − a loop type reactor using mixed oxide fuel − lasted from 1962 to 1966. The reactor had two cooling circuits of 12 MW(th) heat capacity removal each. The first criticality was reached on the 28th January in 1967. The initial power of the reactor was 20 MW(th), and it was increased to 24 MW(th) in December 1967 and to 40 MW(th) in 1971 after some modifications (FORTISSIMO project).
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57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03
01/67 10/83 40 MWth
RAPSODIE 08/73 12/73 250 MWe
PHENIX 09/85 01/86 02/98
1200 MWe SUPERPHENIX
SPX 2 1500 MWe EFR
1st connection to the grid Final shutdown 1st criticality
Studies & design Construction Operation Decommissioning
FIG. 3.4. French fast reactor program. RAPSODIE was an outstanding irradiation tool (max. Burn-up 27 %), allowing demonstration of oxide fuel capabilities and an initial screening of core structural material. The operation of RAPSODIE was excellent from 1967 to 1978 until the discovery in October 1978 of a very small sodium leak (estimated at 10g/year) in an area surrounding the primary vessel. The exact localization and repair of the leak was estimated too long, too costly and too uncertain. Later in 1982 a larger nitrogen leak (estimated at 0.5 cm2) appeared across the double wall of the reactor. It was decided to close the reactor because the double containment was no longer assured. Moreover, RAPSODIE had reached its objectives and at that time, PHENIX reactor had proved to be able to ensure all irradiation needs. Finally, the RAPSODIE reactor was shut down in April 1983, after several end-of-life tests. In particular, experiments simulating hypothetical accidental situations like the loss of heat sink without scram (ULOHS) and the loss of electric power without scram and without back up of pumps (ULOF) were performed. PHENIX The prototype fast reactor, PHENIX, went critical in 1973 and reached full power in 1974, after a short construction period (1968−1973), without difficulties. PHENIX is a pool type reactor. The primary circuit is enclosed in the main vessel, with 11.8 m diameter, and filled with 800 t of sodium. It contains three primary mechanical pumps and six intermediate heat exchangers (IHX) connected with three independent secondary loops. Each secondary loop is connected to a steam generator (SG) consisting of three stages: evaporator, super heater and re-heater, with twelve modules each. Between 1974 and 1990, the plant was used for experimental irradiation and for electricity production (250 MW(e) electrical output) with a very satisfactory operation rate for a prototype installation (51 cycles, 61 % average load factor and more than 20 billion kWh produced). During these 16 years of operation, mainly two periods of unsettled operation (with shut- down or at 2/3 of nominal power) occurred in order to repair and modify IHX (1976−1978) and SGs (1982−1983). In addition some very limited sodium fires occurred due to small
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sodium leaks (few kg) in the secondary circuit. Safety analyses of these incidents showed that they could be seen as a possible common mode failure of the secondary circuits. For future reactors, an effective separation of the circuits has been asked for by safety criteria. In 1989 and 1990, four emergency shut downs were activated by negative reactivity signals, analysis of the origin and research into the potential consequences required a long development which couldn’t find a definite conclusion. Nevertheless a complete review of different scenarios of reactivity change was performed and renewed for SUPERPHENIX (similar scenario could not jeopardize the safety). In the mid-nineties, the role of the reactor changed: it was to be used as an irradiation tool acting as a support tool for the R&D transmutation programme of the CEA within the framework of the 1991 French law concerning long-lived radioactive waste management. As the initial lifetime of the reactor was 20 years, the reactor should have been shut down in 1994. This new objective required an extension of the planned reactor lifetime. Thus, a large renovation programme was defined and achieved in February 2003. Full power was achieved again in June 2003. Six operating cycles, covering about five and a half years, are planned to carry out the irradiation experimental programme on minor actinides and long-lived fission products transmutation, and future reactor options. From the beginning, PHENIX appeared as a reactor quite easy to control and operate. As RAPSODIE, it was a very efficient irradiation facility (Fig. 3.5), thanks in particular to an adjacent hot cell allowing intermediate examinations of pins and return of them to the reactor. Records of burn up on oxide and dose on materials (155 dpa) were achieved and contributed to the optimization of fast reactor fuels.
sub-assemblies 60 1975 1980 1985 1990 capsules 40 total 20 number of
experiments 0 1 4 7 1013161922252831343740434649 Cycles
FIG. 3.5. PHENIX fuel experimental program. But also in the fields of physics and sodium technology, the contribution of PHENIX was considerable: as an example, it was the first reactor to demonstrate a breeding capacity of a fast reactor, by recycling Pu coming from its own previous core elements; it demonstrated also, at several occasions, the possibility to remove big components (e.g., intermediate heat exchangers or primary pumps), to clean and repair them, and then to reintroduce them in reactor for further operation. Another major contribution of PHENIX is the evidence of low radioactive releases, confirmed by more than 20 years of operation (Fig. 3.6).
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TBq Fig 1 - Gaseous releases from Phénix
10,00 Authorisation : 400 TBq/year 8,00 6,00 Average : 5,47 TBq 4,00 2,00 0,00
4 6 8 0 2 4 6 8 0 2 4 6 197 197 197 198 198 198 198 198 199 199 199 199
FIG. 3.6. PHENIX gaseous releases. The renovation programme allowed the reactor to perform irradiation tests in support to the CEA transmutation R&D programme. It was based on: • Safety improvements, to take in account the evolution of the safety standards and today’s design and construction rules; • Systematic expert evaluations and inspections of components to estimate their status and relevant damaging mechanisms; • Assessment of the components capacity to continue operation, using feedback experiments. Before the final shutdown of the reactor scheduled in 2009, a large program of “end of life” tests will be performed to get complementary knowledge in the area of neutronics, thermal hydraulics and safety. Specific test will be devoted to obtain information on possible scenarios related to abnormal negative reactivity transients observed in 1989 and 1990.
SUPERPHENIX SUPERPHENIX has been built and operated at Creys-Malville by NERSA, a company grouping together French, Italian, German, Belgian and Dutch utilities. Its construction lasted from 1977 to 1985; full power was reached in 1986, and till the end of 1996 the operation time was 4.5 years and 320 EFPD. Operation assessment Eleven years of operation from the first electricity generation in January 1986 right up to the end of 1996, are summarised in Fig. 3.7: • 53 months, in other words nearly 4 1/2 years of normal operation with periods of electricity generation at different levels of power, scheduled periods of continued operation and obviously test periods − SUPERPHENIX is a prototype − notably during 1986, finally to reach full power, 1200 MW(e) in December 1986; • 54 months, i.e. 4 1/2 years, although in operational state, the plant remained shut down due to administrative procedures under way; • 25 months, a total of a little more than 2 years, the plant has not been in operational state following technical incidents and necessary repairs.
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FIG. 3.7. SUPERPHENIX operation. Main lessons learned from SUPERPHENIX Because of the building and operation of SUPERPHENIX, it is known that the main technological options for a sodium-cooled fast reactor are industrially valid at full scale. Experience acquired thanks to SUPERPHENIX notably concerns the areas of design, technology, operation and safety: fundamental knowledge and computer codes, principles, methods and operation and maintenance procedures, etc.
• Design, computer codes: o Feasibility of a large size reactor, validation of the design options retained for fuel, circuits, etc.; o Knowledge of operational features of a commercial size fast reactor core; confirmation of its stability and its easy monitoring; o Validation for normal and accident situations of neutron codes applied to a large size flat core (and no longer of a small orthogonal-cylindrical size); o Compilation of a great amount of data (thermal, thermal hydraulics, mechanics of thin shell structures, creep, buckling, etc.), validation and improvement of sizing and operational codes; o Improvement in the knowledge of the mechanics of large austenitic steel structures, called upon in the elastic-plastic field (creep, buckling, etc.), drafting of design codes and criteria allowing to study the harmfulness of defects and their possible evolution, and corresponding development of design, and construction codes and rules.
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• Technology: o Lessons learned from operation and incidents (metallurgy, sodium chemistry, specific instrumentation, remote diagnosis and intervention, etc.). o Better knowledge of the physical chemistry of sodium applied to a very large bulk (purification, solubility of metals and gases, corrosion, deposits, etc.) of radionuclides and gas transfers, detection and localisation of clad ruptures, sodium- hydrogen-tritium balance studies, formation of bubbles and void effect, etc. o Prevention of risk associated to sodium: leaks, fire and sodium-water reaction, perfecting and industrial validation of a protection and leak detection system, important modifications, accounting for spray fires, compartmentalisation of large volumes, acoustic detection, etc. o Accumulation of experience, feedback and a data bank concerning operation and reliability of large components and their back-up: primary and secondary pumps, steam generators, heat exchangers, fuel handling machine, etc.
• Control techniques, operating methods: o Development of new techniques for non-destructive testing adapted to the need for inspection of reactor and steam generator structures covered by sodium (ultrasound, eddy currents, etc.). o Perfection of operating and maintenance methods for complicated prototype systems adapted to the rules and methods applied to PWR plants of the French reactor park; particular attention has been given to an ergonomic approach and consideration of the human factor.
• Safety: As far as safety is concerned, the construction, commissioning, operation and analysis of events and incidents have been an invaluable source of knowledge, for which neither calculation nor experimentation can be a substitute (scale effect, environment and industrial practices, etc.). Experience gained and safety studies performed demonstrate that the safety level of SUPERPHENIX is comparable to pressurised water reactors.
EFR The EFR project, launched in a European frame associating utilities and designers, supported by R&D, completed in 1993 the design of an optimised 1500 MW(e) reactor, with a credible evaluation of economic parameters. Following years were devoted to studies of technical improvements of the EFR design. When the abolition of SUPERPHENIX was announced it was decided to close the work on EFR by the end of 1998.
3.2.2. Main achievements in R&D Fuel Fuel behaviour was a first concern. After the choice of mixed oxide as nuclear fuel for fast reactors , the main issue was the ability to reach a high burn up that is necessary to improve economic performances. RAPSODIE and then PHENIX allowed continuous progress in burn- up and doses (Table 3.6), and a screening between candidates for cladding and wrapper materials.
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TABLE 3.6. SOME OUTSTANDING EXPERIMENTAL ACHIEVEMENTS IN RHAPSODIE AND PHENIX Burn up (%): 27 with RAPSODIE standard fuel. 17.5 with PHENIX standard fuel. 7.5 with nitride fuel in PHENIX. Damage on core materials - claddings in 155 in 15/15 Ti SS. PHENIX (dpa) 123 in high Ni alloy. 130 in ferritic-martensitic steel. 91 in ODS ferritic steel. Damage on core materials - wrappers in 155 in martensitic EM10. PHENIX (dpa) 127 in 15/15 Ti SS. 20 3 Absorber in PHENIX 220 10 neutron captured/cm in B4C. Today, demonstration has been made that a burn-up of at least 150 000 MWd/t (150 dpa) can be reached, and further important progress appears quite possible. These achievements are accompanied by a lot of knowledge on failed fuel behaviour, on transient conditions, and contributed to developing and validating computer codes.
Sodium The utilisation of sodium at a large industrial scale was another key issue. Out of pile tests in specific facilities provided the first technological information, then came the feedback from reactors; then control of sodium quality, operation of circuits with all kind of instrumentation, knowledge of radioactive contamination transfer, and procedures to clean and decontaminate transfer components were mastered. Efficient prevention against sodium-water reaction (in steam generators) does now exist. Substantial improvements have been gained in the key problem of inspection above or under sodium in reactor vessels.
Core physics The success of fast reactor development depended also on the capability to control large cores. Thanks to research work using zero power critical facilities, among which the most noticeable is MASURCA, analytical tools were built, the last of which (ERANOS) became operational in 1997, and validated by a comprehensive series of representative tests. Using these tools for core design the operations of SUPERPHENIX finally showed no neutron specific problem.
Safety It was important to achieve for these reactors the same level of safety as for other types. Both experimental and theoretical studies were conducted, leading to a continuously improved modelling work. The analysis enabled to consider core melting accidents as a reference for design studies. An important role for assessment of phenomena played the experimental tests conducted in the CABRI and SCARABEE facilities on representative fuel elements. Sodium leak prevention and sodium fires studies are another part of activities specific for fast reactors. For this issue also a program of tests was conducted especially in the ESMERALDA facility. Now a correct description of pool spray fires is possible, and the ways to prevent unacceptable consequences are known. There is also now a good experience of steam generator safety problems, both regarding detection of a sodium water reaction, and the knowledge and the control of consequences.
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Structural engineering This short review must also include development of structural mechanics. Fast reactor components typically are operating at high temperatures and are thin, a structure for which phenomena as creep, fatigue or a combination of both, and buckling play a great role. Specific methods were developed based on experiments and integrated into a system of conceptional codes, (called RCC-M-R) which allow now to deal with any design problem. Since some years, PHENIX brought another important contribution, namely, demonstration of the evidence of ageing phenomena in materials, having accumulated about 100 000 hours in service.
CAPRA To conclude this review of main French R&D achievements, some information on the CAPRA programme; initiated in 1992 is presented. The programme demonstrated the flexibility of fast reactors, which using available neutrons can contribute either to breeding, or burning of plutonium, or to destroying minor actinides and long lived fission products. New fuel elements and core designs are explored in the frame of this programme.
A summary of the accumulated knowledge: outcomes of EFR studies Since the launch of the EFR design studies in 1988, a substantial amount of work has been completed and excellent results have been obtained, thanks to the efficient collaboration between all the participants: EFR associates, the utilities and R&D centres. Even though political circumstances have required the EFR project organization (EFRUG) to give up in 1993 the initial objective « to enable construction of a new fast reactor in EUROPE », the task to preserve the most of European knowhow, to maintain expertise and to explore innovative ideas in the fast reactor field has been achieved. One of the major achievements of the EFR programme was, besides its own developments, the considerable amount of R&D work performed, which was used in support of SUPERPHENIX, and its knowledge acquisition programme and the feedback of its operating experience for improving the design of future reactors.
3.2.3. The transition scenario from present PWRs to fast reactors in France
The introduction of 4th generation fast neutron systems in the 21st century satisfies the objectives of making optimum use of natural resources and minimizing production of long- lived radioactive waste. The introduction date depends on the industrial maturity of these systems; based on the current state of knowledge, it is estimated that it could start around 2035. Adaptation of fuel cycle facilities for this 4th generation of reactors should be organized taking into account the need to renew existing fuel cycle plants and to optimize new investments. The studied scenario considers two phases in the renewal of French nuclear reactors: • First a renewal of existing reactors between 2020 and 2050, based on EDF's assumptions with a replacement of the Generation-II 5 PWRs by EPRs (Generation-III+) or Generation-IV systems, at a rate of 2 GW(e)/year of installed power; • Second a renewal of existing reactors from 2080 to 2110, assuming a 60-year life of 3rd or 4th generation reactors.
5 Generation-II, Generation-III and Generation-IV are types of reactors classified in the GIF (Generation IV international forum) project.
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Renewal of fuel cycle plants could be planned to start by 2025. Several Generation II - Generation III / Generation IV transition scenarios are considered during the 21st century: • The starting point of the scenario is the year 2010, and it is supposed that the energy supplied by a 60 GW(e) reactor fleet is constant and equal to 400 TWh(e) per year; • About 91 % of this energy is supplied by UOX fuels and 9 % by MOX fuels at 45 GWd/t mean burnup; • Average burnup of UOX and MOX fuels is increased to 60 GWd/t by 2013; • Until 2020, the UOX spent fuels reprocessing addresses the partitioning of plutonium only; under “just-in-time” Pu flow conditions (processing, manufacturing, and reactor) are extended, thus not all spent fuel assemblies are processed; the produced glasses for final disposal contain fission products, MA, and Pu processing losses (0.1 %); • In a later period, around 2040, Generation-IV fast neutron systems, designed for a grouped recycling of all actinides appear as the best candidate among nuclear systems; they will efficiently use natural uranium and will be environmentally friendly, thus contributing – together with other energy sources – to sustainable development; • This leads to the development of scenarios benefiting from industrial plant renewal (reactors and fuel cycle plants) while optimizing the utilization of existing plants, and from their replacement allowing to implement innovative processes and technologies. The reference scenario considers a grouped recycling of all actinides (U-Pu-MA). It foresees around 2040 both the deployment of fast reactors in the French generating fleet (sodium or gas cooled), and the startup of a new spent fuel treatment plant, superseding the existing plant of La Hague. • The effective deployment of this scenario is linked to the progress sought on the quality of ultimate waste forms (end state), and the technical and economical optimization of a geological repository (the design of which depends on the long term decay heat). Main trends for reactors and associated fuel cycles are summarized in Fig. 3.8. For simplification and modeling purposes, the reference scenario for the transition phase is considered as follows: 2020–2035: Start of renewal of 50 % of the fleet with EPR reactors; this renewal relates to the end of the service life of the first PWR plants introduced in 1975/1985 and is carried out, depending on EDF prospects, at a rate of 2 GW(e) per year. 2035–2040: Start of renewal of the remaining 50 % of reactors of the previous generation by Generation-IV fast neutron systems and implementation of advanced reprocessing of spent MOX fuel for plutonium and minor actinides recycling. 2080: Start of renewal of the EPRs which were first introduced in 2020 by Generation-IV fast reactors.
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1975 2000 2025 2050 2075 Reactors Lifetime extension Gen IV Operating Fleet EPR
Source : EDF, ENC 2002 Cycle
U (Udep, URT) Recycling in Gen IV Gen IV Pu (recycling as MOX fuel) Cycle M.A. ---> glas or recycling M.A. + F.P. ---> Glas waste Storage F.P. ---> Glas
FIG. 3.8. Evolution of the French nuclear reactor fleet and fuel cycle facilities. For modeling purposes, the following deployment scheme (Fig. 3.9), was assumed for the period 2000–2100.
70 GWe
60
GEN 2 50 GEN 3 (EPR) GEN 4 (FR) Total 40
30
20
10 years
0 2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
KoreaFIG. 3.9. Simulated transient French scenario.
3.2.4. Main trends of R&D efforts for fast reactor development in France
The R&D strategy for the development of future nuclear systems was approved by the French government in March 2005 and is focused on three components: • The development of fast reactors with a closed fuel cycle; • Nuclear hydrogen production by the development of a very high temperature reactor (VHTR) with a water splitting process and a very high temperature process-heat supply to the industry; and • Innovations for light water cooled reactors regarding fuel and reactor systems. The development of fast reactors with a closed fuel cycle includes two tracks:
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• Sodium cooled fast reactors with the objective of starting the operation of a prototype by 2020; • Gas cooled fast reactors with the objective for an experimental reactor in an European framework. These tracks are associated with the development of new processes for spent fuel treatment and recycling. The final objective is to reach industrial maturity for commercial deployment around 2040. Sodium cooled fast reactors The sodium cooled fast reactor has the credit of a very high experience gained in the world especially in France, Russian Federation and Japan. It is the most mature of all fast reactor concepts. It was selected as one of the Generation-IV promising concepts despite some limitations: • Sodium drawbacks; • Limited temperature for some applications; • High investment cost; and • Some safety issues. The R&D to solve the difficult points (limitations) of these concept – primarily a needed investment higher compared to PWRs – includes: • Eliminate the intermediate circuit; • Design and evaluate new concepts; • Improve in-service inspection and repair; and • Enhance safety be reducing a re-criticality risk for severe accidents. In line with the objective of waste reduction developing fuel allowing grouped actinides recycling is a common R&D direction for fast reactors, including a re-fabrication process. In January 2006 the president announced that the French atomic energy commission (CEA) was to embark upon designing a prototype Generation-IV reactor to be operating in 2020, moving forward the timeline for this project by some five years. France has been pursuing three Generation-IV technologies: a gas-cooled fast reactor, a sodium-cooled fast reactor, and a very high temperature reactor (gas-cooled). Gas cooled fast reactors Gas cooled fast reactors (Fig. 3.10) have some interesting features: • Neutronics transparency; • No phase change (no cliff edge effects); • Chemical inertness; • Optical transparency.
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Figure 3.10. Gas cooled fast reactor.
Common technologies for materials and components can be developed in line with the VHTR dedicated to hydrogen production. But anyway there is a need for a large R&D program giving access to both “fast spectrum” and “high temperature”. Gas cooled fast reactors’ innovative fuels are characterized by: • Robustness and refractory; • High level of fission products confinement; • Increased resistance to severe accidents. The design objectives of this concept are: • Self-sustainable core; • Robust safety approach; • Attractive power density (~100 MW/m3).
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3.3. INDIA
3.3.1. Planning of capacities and structure of nuclear power sector
India has a vision of a three stage nuclear power program (Fig. 3.11). In the first stage, pressurized heavy water reactors (PHWR) are deployed which will use natural uranium as fuel. Indian PHWR is currently in a commercial phase. Sixteen PHWRs and two boiling water reactors (BWR) with a total generation capacity of 3850 MW(e) are in operation. Six reactors are under construction, which includes two VVERs. Liquid sodium cooled fast breeder reactors (FBR) form the second stage of the nuclear power program. Plutonium (Pu) generated from PHWRs would be used as fuel within a closed fuel cycle program in fast breeder reactors. Subsequently, thorium resources will be utilized in third stage reactors with Pu and U233 as fuel coming from FBR.
FIG. 3.11. A three stage program of nuclear power development in India.
Apart from thermal reactors, a 500 MW(e) FBR is, presently, also under construction. The share of FBR in energy production by 2020 is expected to be ~3 GW(e) in an overall nuclear share of 29 GW(e) forming a part of the total projected capacity of 417 GW(e). However, by 2050 the total nuclear contribution will be 275 GW(e) (contribution of FBR would be ~260 GW(e)) as part of a total projected capacity of 1344 GW(e) (Fig. 3.12).
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FIG. 3.12. Projected installed power capacity in India.
This significant increase in capacity until 2050 can be achieved only by choosing a closed fuel cycle with a high breeding system. It is therefore planned to start a comprehensive program on the development of advanced oxide and U-Pu-Zr metal alloy fuels in breeder reactors and associated fuel cycle technologies in a systematic and prioritized fashion. Emphasis would be placed on innovation, scientific breakthrough, development of quality human resources and robust technologies through synergy between academic-research and industry. FBR with metallic fuel is the chosen option of shifting to a high breeding path by 2020. Thorium would be irradiated in the second stage FBRs with the objective of generating 233U which will form the fuel for the third stage reactors. Towards this direction, advanced heavy water reactors (AHWR) are planned to demonstrate the technology of innovative thorium based reactors and fuel cycle; the AHWR design is under regulatory review presently. Clearance for construction of AHWR is expected by 2007.
3.3.2. Brief history and current status of fast reactor technology development
India started its FBR program by constructing an experimental fast breeder test reactor (FBTR) at Kalpakkam (Fig. 3.13). The FBTR was commissioned in 1985 with an unique plutonium rich carbide fuel. FBTR is a 40 MW(th)/13.2 MW(e), sodium cooled, loop type fast reactor with two primary and two secondary sodium loops. Each secondary loop has two once-through serpentine-type steam generators (SG). A turbine generator (TG) and a 100 % steam dump condenser (DC) to facilitate reactor operation without TG are also provided. The first criticality was achieved on 18th of October, 1985, with a small core of twenty-two fuel subassemblies (SA) of MK-I composition (70 % PuC – 30 % UC), with a design power of 10.6 MW(th) and a peak linear heat rating (LHR) of 250 W/cm for a twenty-nine SA core. Progressively, the core was expanded by adding SA at peripheral locations. During the increase of the core size and hence the reactor power, carbide fuel of MK-II composition (55 % PuC – 45 % UC) was inserted into peripheral locations in 1996. TG was synchronized to the grid for the first time in July 1997.
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FIG. 3.13. Integral layout of Indian FBTR.
The reactor has so far been operated up to a power level of 17.4 MW(th) by raising the LHR of MK-I fuel to 400 W/cm in 2002. So far thirteen irradiation campaigns have been completed and the 14th irradiation campaign is in progress. FBTR operates with a high availability factor (> 90 %) in irradiation campaigns. The present core has twenty eight MK-I and thirteen MK- II driver fuel SA and one test SA simulating the fuel composition of MOX fuel for the Indian prototype fast breeder reactor (PFBR). Based on the experiences gained with international FBRs, including the FBTR, and on a comprehensive research and technology development program in collaboration with a large number of academic and industrial organizations, a 500 MW(e) capacity prototype fast breeder reactor (PFBR) is being established (Fig. 3.14).
Fig. 3.14. Indian PFBR reactor under construction (2007).
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It is a pool type reactor with two primary and two secondary loops with four steam generators per loop. It will use mixed oxide fuel with two enrichment zones (21 % and 28 % PuO2). 20 % cold worked D9 material (15 % Cr-15 % Ni with nano-size TiC precipitates) will be the core structural material. SS316 LN is the major structural material for the sodium components and modified 9 Cr-1 Mo is the material for steam generators. The sodium temperatures are 820 K / 670 K for hot and cold pool, respectively. The plant has been designed for a life of 40 years with at least a 75 % load factor to be successively enhanced to 85 %. The design and development activities of the PFBR are nearly completed. The construction of PFBR in Kalpakkam has been started and the reactor should be commissioned before September 2010. All the nuclear steam supply system components are being manufactured by Indian industries, based on domestic technology development, including in-sodium testing of relevant components over the last two decades. Fig. 3.15 shows a schematic of the PFBR reactor assembly.
FIG. 3.15. Schematic of PFBR reactor assembly.
3.3.1. Incentives and boundary conditions for future deployment of INS
India is a large country with a high density of population and is on a fast growth path. During 1981−2000, an impressive average domestic product (GDP) growth rate of about 6 % /a has been achieved. Continued growth of Indian economy is forecasted and development activities are planned based on a well-conceived road map and a clear vision. Like elsewhere, the energy and electricity growth in India are also closely linked to growth of the economy. The average growth rate of energy since political independence is about 9.54 % /a which is indeed impressive. The current GDP growth rate is 7−8 %. In order to achieve the planned faster growth rate of 10 % GDP per annum, the government of India is exploring the exploitation of all energy resources in a comprehensive and complementary fashion.
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A survey by the Indian department of atomic energy forecasts an electricity growth rate of 6.3 % /a until the year 2020 and about 4.0 % /a from 2020 to 2050. The electricity potential of various energy resources in India (in GW(e)a) are: 7614 for coal), 5833 for hydrocarbons, 328 for uranium in PHWR, 42 231 for uranium in FBR, 155 502 for thorium in breeders, 69 for hydro-renewables and 33 for non-conventional-renewables (annual). The reference scenario projects cumulative usage of main resources by 2050 to be 943 EJ from coal, 912 EJ from hydrocarbons, 246 EJ from nuclear, 212 EJ from hydro-renewables and 72 EJ from non- conventional renewables (Fig. 3.16). According to this projection, the presently known extractable coal reserves would be exhausted by 2050.
FIG. 3.16. Potential of primary energy resource and scenario of energy generation in India.
Based on the known technologies and expected energy consumption growth pattern, it is imperative that the growing energy demand should be met through fast breeder reactors (FBR) and thorium fuelled reactors, in combination with other energy resources, taking into account sustainability and environmental issues.
3.3.4. Scenarios for deployment of a closed nuclear fuel cycle with FBR
India has always adopted a closed fuel cycle as an important component of its nuclear energy planning, thereby considering the advantages of a closed fuel cycle. It has recognized that an effective utilization of its uranium resources as well as plutonium generated in PHWRs can only take place by adopting a closed nuclear fuel cycle. In line with this approach, a fuel cycle facility is being constructed at the location of the PFBR in order to close the fuel cycle of PFBR at the earliest. A similar approach will also be adopted for future FBRs as well.
The time frame for deployment of FBRs with a closed fuel cycle is presented below: • 2010: Commissioning of the PFBR; • 2012: Commissioning of a fast reactor fuel cycle facility for the PFBR;
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• 2012: Start of construction of four FBR-500 (two twin units); • 2018: Commissioning of four FBR-500; • 2018: Commissioning of a fast reactor fuel cycle facility for four FBR-500; • 2018: Start of construction of two 1000 MW(e) oxide fuelled FBR (FBR-1000); • 2020: Change of core of one FBR-500 to metallic fuel; • 2024: Commissioning of two FBR-1000 reactors; • Beyond 2025: Series of 1000 MW(e) metallic fuelled FBRs with a matching fuel cycle.
3.3.5. Fast reactor design approach with a closed fuel cycle
From the point of view of economy and experience gained / to be gained in construction and operation, it is planned to go in for a larger fast reactor (500, 1000 MW(e)), rather than smaller sized modular type reactors. Sodium will be used as the coolant. While oxide fuel will be inserted into the first four units of 500 MW(e) reactors, the reactors of 1000 MW(e) to be constructed subsequently will use a metallic alloy as fuel material. The emphasis of research would be developing metallic fuels which can deliver high breeding ratios (minimum zirconium content in alloy) and which can operate to a high burn-up and at high temperatures. A series of reactors would be built with innovative safety features and improved economy with a provision to change over from oxide to metal fuel. The reprocessing of oxide fuel used in the first twin units will be carried out using the well- developed PUREX process. However, pyroelectrochemical processes will be established for reprocessing the metal alloy fuels from subsequent fast reactors. Several innovative features would be introduced to enable reprocessing of fuel with a high burn-up (> 100 000 MWd/t) and short cooling time (180−240 days). It is planned to have no delay in reprocessing of the spent fuel after it has been cooled in the reactor pool between successive fuel handling campaigns. This would reduce the out-of-pile inventory of Pu and contribute to enhancing the power capacity expeditiously. In addition to these measures, emphasis will be laid on reducing waste generation and on-line monitoring of actinides and fission products to reduce losses to waste streams, and enable better process control and nuclear material control. The means to achieve improved economy and safety have been identified systematically. A few of them are highlighted below. Enhanced plant life with higher load factor For FBR-500, a sixty years design life with an 85 % capacity factor is envisaged. Further, based on a critical re-investigation of a number of design basis events (DBE) that can reduce fatigue damage and by a choice of better materials with high creep strength, even a 100 years plant life can be thought of. Higher plant efficiency It is planned to choose plant parameters in a way that steam temperature can reach 803−813 K, to match the steam conditions of fossil fuelled power plants, leading to an increased plant efficiency in association with reduced thermal pollution. The resultant plant efficiency is expected to be greater than 42 %. High burnup Higher burn up to 200 GWd/t would be considered as a target. Choice of advanced materials and improvement of present materials through characterization, design of materials, modeling
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and performance studies constitute the primary mechanism of achieving economy through higher burnup. High breeding ratio With the optimization of parameters such as composition, fabrication methodology, fuel design options, burnup and breeding ratio, the breeding ratio targeted would be around 1.6 for metallic fuel and 1.16 oxide fuel. Advanced structural materials Use of SS304 LN instead of SS316 LN, especially for the cold pool components and ferritic steel especially modified 9 Cr-1Mo for the entire secondary sodium circuits are promising choices. Improved primary circuit components With an optimization of space in the two loop design, the primary sodium purification circuit will be incorporated within the reactor assembly. Optimized main vessel dimension By integration of control plugs with small rotatable plugs and elimination of the main vessel cooling circuit, the main vessel diameter will be reduced significantly. Thick plate top shield A thick plate concept instead of a box type design will be adopted for the top shield. This concept can also reduce the thickness of the top shield significantly and hence, the overall height of components passing through the top shield. Further, the thick plate top shield has significant functional advantages by resulting in a reduced height and smaller annular gaps between penetrations. The latter advantage leads to reduced radiation streaming, complimentary shielding and sodium ejection during core disruptive accidents (CDA). Anchored safety vessel In view of significant economic advantages, an anchored safety vessel concept is considered in which the safety vessel and liner are integrated. Minimum number of steam generators With the choice of modified 9 Cr-1Mo material, it is planned to have only 2/3 SG modules per loop which in turn reduces the capital cost. Compact plant layout A compact layout can be achieved by an optimum piping layout (placing components and inter-connecting pipelines predominantly in a vertical plane with minimum horizontal spread) and by improved layout of civil structures with a possible minimum building size. This also results in increased structural reliability due to less number of welds. State-of-the-art seismic design A rational approach of seismic design in selecting site parameters (peak ground acceleration and spectrum shape), damping values (FBR structures have higher damping values due to the large size of their components, higher insulation thickness, presence of a large number of snubbers, etc.), careful categorization (Category 1 and Category 2 items need to be re- investigated in compliance with safety) and classification (for reactor assembly components, OBE at level C in line with international practices) will reduce the cost significantly. Rigorous seismic analysis also helps to reduce the cost. Apart from this, a base isolation
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concept with the use of seismic isolators is an ideal solution to achieve economy and standardization. In the following salient features to achieve enhanced safety are presented: Reliable shutdown systems Two independent diverse shutdown mechanism having distinctive designs incorporating passive features are necessary to meet the specified reliability requirements. Additionally, optical coupling between them further enhances the reliability of action on demand. Passive featured decay heat removal system Diversity and redundancy are built into the safety grade decay heat removal systems in the form of variant designs with sufficient margins and passive features. Further, an in-vessel purification concept for radioactive primary coolant enhances the safety of the reactor. Rationalization of DBE Rationalization of the design basis earthquake (DBE) categorization and long term event progression can be evaluated by detailed probabilistic safety analysis (PSA). This, apart from raising the confidence in plant dynamics studies, will result in a reduced number of anticipated DBE. This also contributes to enhanced plant life. Elimination of CDA By including provisions to enhance the inherent safety with a modified design approach, and considering further aspects which enhance inherent safety, such as a heterogeneous core concept and an extended primary flow coast down following an off-site power failure, safety measures would render an anticipated transient without SCRAM (ATWS) impossible, by which the core disruptive accident (CDA) can be eliminated in the design. These measures would minimize the requirement of several advanced analytical and experimental studies and design provisions. ISI and repair of reactor internals Emphasis will be given on an improved in service inspection (ISI) and repair for reactor internals to ensure a reliable operation with an enhanced plant life. This also provides economic advantage by way of eliminating undue and needless conservatism in the design. Under sodium viewing using ultrasonic means is becoming more and more attractive. By taking advantage of such techniques coupled with recent advances in robotics, inspection and repair of reactor internals can be accomplished more easily adding significantly to plant economics. Detailed analysis with realistic material constitutive models encompassing all life limiting damage mechanisms coupled with advanced inspection techniques have the possibility to contribute to economy and safety. In the following salient features of the fuel cycle are presented: Fuel fabrication Fuel fabrication will be carried out using a powder metallurgical route (milling and blending of oxides followed by pelletisation and sintering); dose to workers would be reduced by introducing remote inspection techniques; dry scrap recovery techniques would be used for recycle of rejects in order to reduce aqueous waste production. Sol-gel based fuel fabrication methods enable remotisation and automation due to the fluid like behaviour of the microspheres resulting in a reduction of dose (man-rem) to workers in the absence of powder handling. Sol-gel microsphere pelletisation offers the advantages of sol-gel based methods as well as the rich experience on pellet fuel irradiation. Hence efforts
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will be made to incorporate sol-gel microsphere pelletisation in the fabrication flow sheet. An economic and safe flow sheet will be developed for metallic fuel fabrication. Fuel reprocessing An integrated flow sheet for the reprocessing and waste management of FBR fuels is shown in Fig. 3.17.
INTEGRATED FLOWSHEET FOR ADVANCED FBR FUEL REPROCESSING
FBR/ADS MOX Fuel Based P&T FBR MOX Fuel FF
Spent Fuel Np Tc U+Pu U Disassembly & TBP/Amide Based HAF Dissolution Solvent Extraction in CE 129I as NaI
HAW Actinides Actinide Partitioning and Group Separation Steps
Pd, Sr & Cs Deep Underground Geological Hulls, SW and Isolation & Disposal & Monitoring Insolubles Societal Use Ln/ Other FPs
FIG. 3.17. Integrated flow sheet for the reprocessing and waste management of FBR fuels. The salient features of a reprocessing flow sheet would be: • Fuel cooling time: In line with the time between two successive fuel handling campaigns; • Fuel burn-up: Successively enhancing the capability from 100 to 150 to 200 GWd/t; • Fuel reprocessing time: In line with the time between two successive fuel handling campaigns with no additional cooling time for the fuel; • Process: PUREX process employing long chain amide as the extractant. This would provide following benefits: • Low solvent solubility in the aqueous phase; • U/Pu separation without use of reducing agents; • Co-precipitation of U and Pu; • Incineration of used solvent without leaving any residue; • Non-salt forming agents can be used for all operations. Co-precipitation of U and Pu would enhance nuclear material accounting, control and physical protection. Pyrochemical and aqueous reprocessing techniques will be examined and developed for metallic fuel processing. Waste management The salient features of waste management philosophy will include the complete recovery of minor actinides from high level waste, rendering it more amenable for disposal in near ground
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repository after immobilization. Significant reduction in the volume of liquid waste will be achieved by improvements in the flow sheet. Integrated fuel cycle facility An integrated fuel cycle facility comprising fabrication, reprocessing and waste management plants will be located at the reactor site itself which will be economical, safe and also reduce problems of transport of active material through public territory.
3.4. JAPAN
3.4.1. Brief history and current status of fast reactor technology development
Japan’s fast reactor developments took a step-by-step approach from an experimental stage. An experimental fast reactor Joyo (Mk-I: 50 MW(th), Mk-II: 100 MW(th), Mk-III: 140 MW(th)) has been in operation successfully for about 30 years since the initial criticality in April 1977. It has been used for several objectives, such as confirmation of fundamental breeding characteristics, mastering fast reactor operation, inspection technology and fuel irradiation technology. Furthermore, it has contributed to the development of fuels and materials for the prototype fast reactor MONJU and future commercial reactors, and it also has been utilized by outside organizations such as universities. The Mk-III program, which consisted of the renovation of Joyo in order to increase irradiation capability, was completed on November 27th, 2003, and irradiation tests were re-started in 2004 in order to develop advanced fuels and materials. MONJU is a prototype power-generating fast reactor designed to have an output of 280 MW(e) (714 MW(th)), fueled with mixed oxides of plutonium and uranium, and cooled by 3 sodium loops. The development objectives of MONJU are; to demonstrate performance, reliability and safety of a power-generating fast reactor and to establish a sodium-handling technology through collection of experiences in design, fabrication, construction, operation and maintenance of the plant. MONJU successfully achieved its first criticality in April 1994, and supplied electricity to the grid initially in August 1995. However, the pre-operational test of the plant was interrupted by a sodium leak accident in the secondary heat transport system in December 1995 during a 40 % power operation. After carrying out a cause investigation, a comprehensive safety review and the necessary licensing procedure, a permit was issued by the regulator. Approval for modification start of the MONJU plant was given by the local government of Fukui prefecture on February 7, 2005. Criticality after the accident is expected to be achieved again in 2009. A feasibility study on commercialized fast reactors including related nuclear fuel cycle systems (hereafter called F/S) has been started in Japanese fiscal year 1999 by a Japanese joint project team of Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC) as the representative of the nine electric utilities, in cooperation with the Central Research Institute of Electric Power Industry (CRIEPI). This project aims at elucidating prominent fast reactor cycle systems that will respond to various needs of society in the future, together with economic competitiveness of future electricity supply systems. Challenging technology goals for fast reactor cycle systems investigated in F/S were defined in five development targets, namely, safety, economic competitiveness, reduction of environmental burden, efficient utilization of nuclear fuel resources, and enhancement of nuclear non-proliferation (Fig. 3.18).
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3.4.2. Incentives and boundary conditions for the future deployment of INS CNFC−FR
Contribution to a stable energy supply and measures against global warming Japan’s self-sufficiency rate of energy supply (excluding nuclear energy) is a mere 4 %, the lowest among developed countries, and Japan imports most energy resources from foreign countries.
Safety • Risks caused by introduction of FR cycle should be small compared with risks that already exist in
society.
Economic Competitiveness • Achieve power generation cost comparable to that of future LWR cycle systems and other energy resources. • Ensure cost competitiveness in the global market.
Reduction of Environmental Burden • Reduce the amount of radioactive waste generated in the course of plant operation and maintenance as well as decommissioning. • Reduce the radiotoxicity of radioactive waste by means of burning or transmuting long lived radioactive nuclides (TRU and LLFP ).
Efficient Utilization of Nuclear Fuel Resources • Produce sustainable nuclear fuel. • Respond to diverse needs for future energy resources.
Enhancement of Nuclear Non-Proliferation • Reduce burden of nuclear PP and safeguards (no pure plutonium in any FR cycle process & increase radioactivity of fuel materials). • Effectively operate non-proliferation system (remote process and monitoring system, etc.).
FIG. 3.18. Five development targets for the Japanese feasibility study. Nearly half of the primary energy supply is oil, of which 87 % comes from Middle Eastern countries. On a global scale, as demand for oil will drastically increase due to economic and population growth, mainly, in developing countries, it is predicted that the relation between supply and demand of fossil fuels will become tight, followed by consequent price hikes, and therefore the world may face intensified competition for the acquisition of fossil fuel sources. As it is difficult for Japan to receive directly electricity from neighboring countries − considering the fact that Japan is an isolated island country − it is indispensable for Japan to evolve further into an energy-saving society on the demand side and to ensure stable and reliable energy supply by diversifying import sources on the supply side. Global warming, which will be furthered by the impact of increased energy demand in the world, is one of the most critical environmental issues looming on the horizon, and is a possible threat to human existence. The Kyoto Protocol, adopted as the first step toward long- term and continuous reduction of greenhouse gas emission, entered into force in February 2005, and Japan faces the task of reducing such greenhouse gas emissions to a level of 6 % below its 1990 emission level during the first commitment period, i.e. between 2008 and
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2012. Therefore, Japan needs to maximize the use of energy sources that emit less carbon dioxide, the major greenhouse gas, in addition to making maximum efforts to save energy. Specifically, besides continuing a maximum effort to save energy, it is important to shift energy supply to fuels that emit less greenhouse gases or non-fossil fuels. New energies such as solar and wind power are non-fossil fuels in this category. However, their energy density is low, and the issues of economic efficiency and stable supply need to be addressed at present. On the other hand, nuclear power generation is recommended as a promising mean to contribute to long-term energy supply and as a measure against global warming due to the following characteristics: uranium resources are located in politically stable countries; carbon dioxide is not emitted in the process of power generation; the total amount of carbon dioxide emitted during the lifecycle is about the same as solar and wind power and, when compared with natural gas-fired power generation − which emits less carbon dioxide than oil and coal fired power generations − emission is even lower; radioactive wastes can be disposed of without significantly affecting the environment of people; nuclear power generation further improves the stable supply of energy by recycling nuclear fuels; and commercialization of a fast-breeder reactor (FBR) cycle can drastically improve efficiency of nuclear fuel resource utilization. Therefore, it is rational for Japan to adopt a so-called optimum energy supply mix, in other words, to maximize the use of both new energies sources and nuclear power, taking advantage of characteristics of both. Needless to say that policies will be periodically reviewed based on assessment from various aspects in response to changing technologies as research and development of energy technologies are performed on a continuous basis. Therefore, in order to continuously position nuclear power generation as a key player in energy sources to support the sustainable development of society, all concerned parties need to commit themselves to constantly improving safety, economic efficiency, environmental adaptability, and nuclear proliferation resistance of the technologies. Significance of the fast reactor cycle system deployment [12] Fig. 3.19 shows an estimation of cumulative natural uranium demand in the world based on the evaluation of long-term energy supply and results from the world energy conference (WEC) in 1998. The continuous use of natural uranium in a LWR once-through scheme may lead to depletion of uranium resources. The cumulative uranium demand will exceed the amount of known conventional uranium resources (KCR) around 2040 (Fig. 3.19) in both the “middle course scenario” (Case B) and the “ecologically driven scenario” (Case C2). In the latter half of the twenty-first century, the cumulative uranium demand may exceed the amount of “ultimate resources” (KCR + estimated additional resources).
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FIG. 3.19. Cumulative natural uranium demand in the world [12].
In Japan (Fig. 3.20) the continuous use of a LWR once-through scheme will bring an increase of cumulative natural uranium demand, and the amount will exceed 5 % of the ultimate resources to be consumed in Japan around 2070. In contrast, the fast reactor cycle system will drastically suppress cumulative natural uranium demand in case of a full-scale introduction in 2050.
*1:10% of total conventional uranium resources *2: 5% of total conventional uranium resources 2.5 FBR（Sodium-MOX) h ug LWR(Pu recycling) ro 2.0 -th LWR(once-through) ce on R W *1 L R 1.48 million ton W 1.5 n L g i lin yc ec u r 1.0 P 0.74 million ton*2 FBR cycle system 0.5
0.0 Accumulative uranium demand (million tonU) 2000 2050 2100 2150 2200 Year
FIF. 3.20. Cumulative natural uranium demand in Japan.
Fig. 3.21 shows a comparison of the reduction of the long-term radio-toxicity of high level waste (HLW) between a fast reactor cycle (FR) scheme and a light water reactor (LWR) scheme.
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FIG. 3.21. Radiotoxicity reduction of high level wastes by changing from a LWR system to a fast reactor(FR) system.
The dotted line in Fig. 3.21 indicates a radio-toxicity level of natural uranium ore needed as fuel supply for a million kW(e) power plant per year. The radio-toxicity level for a fast reactor cycle scheme is approximately 1/240 of a LWR once-through scheme on a natural barrier warranty of about 1000 years. The fast reactor cycle system may greatly contribute to the establishment of Japan’s energy security and may help drastically reduce the radio-toxicity level of HLWs. Thus, the commercialization of a fast reactor cycle system is extremely important for a sustainable use of nuclear power.
3.4.3. National planning of capacities and structure of energy sector with determination of share of nuclear power and fast reactor
The energy supply and demand subcommittee in the advisory committee for natural resources and energy of the ministry of economy, trade and industry (METI) of Japan reported a "long- term outlook for energy supply and demand" in March 2005. According to a reference case based on advisory committee’s investigation, Japanese energy consumption grew strongly in the past, but within the next 30 years it will peak in 2021 and then will decrease to 425 million kl equivalent to oil because of a population decline and economic and social structural transformations. Final energy consumption in 2000 was 413 million kl. On the other hand, electricity demand is expected to continue to increase steadily from 940 TWh in 2000 to 1220 TWh in 2030. Though the share of oil in total primary energy supply will fall gradually from 47 % in 2000 to 38 % in 2030, it will remain an important energy resource for Japanese energy supply in 2030. It is anticipated that the share of nuclear and natural gas in total energy supply will rise to 38 % and 30 % b 2030, respectively. Recently, METI announced a “new national energy strategy” in May 2006. Objectives of this strategy are the “establishment of energy security measures”, “establishment of the foundation for sustainable development”, and “commitment to assist Asian and world nations in addressing energy problems”. Some numeric targets are set in order to establish energy security measures:
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• At least another 30 % improvement of energy efficiency should be attained by 2030. • The ratio of oil to total supply of primary energy should be reduced below 40 % by 2030. • The percentage of oil dependence in the transport sector should be reduced to around 80 % by 2030. • The ratio of nuclear power to total power production should be maintained or increased to a level of 30 to 40 % or more by 2030 or later. • Volume of oil exploration and development by Japanese companies should be increased to around 40 % by 2030. Japan imports most of energy resources (approximately 96 %) from overseas. Thus, Japanese energy supply structure is fragile. To improve this situation, Japan has developed nuclear power for the last fifty years based on the principle of peaceful use, and now 55 nuclear power plants (NPP) (28 BWR, 4 ABWR, and 23 PWR) are in commercial operation with a total installed capacity of about 50 GW(e) in June 2006; one ABWR and one PWR are under construction and eleven NPPs are planned. Nuclear power provides an extraordinary stable energy supply and generates 16 % of primary energy supply in Japan. Nuclear power supplies one-third of electricity and the rate of energy resource import is improved to 80 % (from 96 %) if nuclear power is considered as a domestic energy resource. Nuclear power generation is an important main power supply system in Japan and contributes to stabilization of domestic total energy supply and limitation of greenhouse gas emissions. In addition, Japan has promoted the development of a closed nuclear fuel cycle to enhance the efficient use of uranium resources and to reduce high-level radioactive wastes (HLW) as a national policy. Progress has been achieved in some fields including uranium enrichment and nuclear waste management. A 1050 t-SWU enrichment plant and a low-level radioactive waste disposal facility are in operation. The Rokkasho reprocessing plant with an annual throughput of 800 t(HM) has started the uranium tests and its commercial operation is scheduled to begin in 2008. The construction of a mixed-oxide (MOX) fuel fabrication plant is also in progress at the Rokkasho site. Plutonium extracted from the reprocessing of spent fuel will be recycled into light water reactors as MOX fuel. The legal framework of the disposal of HLW was promulgated in 2000. Potential sites are now being surveyed in accordance with the law, and construction and operation of additional nuclear facilities are planned to commence by late 2030. The Atomic Energy Commission of Japan issued a “framework for nuclear energy policy” on October 11th, 2005. This outline was approved by the cabinet on October 14th, 2005. In this framework, following items are noteworthy: • Current level of the share of nuclear power generation should be kept or increase to 30 % to 40 % of total electricity generation even after 2030; • Advanced models of the current LWR should be prepared for replacement of existing plants, which will start around 2030; • Commercial use of fast reactors (FR) should be achieved before or after 2050 meeting the necessary preconditions; • Traditional basic policy that spent fuels are reprocessed and recovered plutonium and uranium are effectively used should be kept.
3.4.4. Optimistic and moderate scenario for development of a closed nuclear cycle with fast reactors
In the present transition scenario study, fast reactors (FR) are assumed to be deployed for commercial use continuously after 2050. In FR deployment scenarios, Pu recycling in LWR
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will be terminated before FR introduction, and then Pu and MA recovered in the future LWR reprocessing plant around 2050 will be reused in a FR cycle. MOX spent fuels discharged from LWRs are treated together with FR MOX fuels in FR reprocessing plants, and Pu and MA recovered from LWR MOX spent fuels will also be reused in a FR cycle in sodium- cooled reactors with MOX fuel. In addition, scenario studies corresponding to a LWR once- through and to Pu multi-recycling in LWRs were carried out for reference to be compared with a FR cycle deployment scenario. According to the "long-term outlook for energy supply and demand" aforementioned, the nuclear power generation capacity in 2030 is expected to increase to 58 GW(e) from presently 50 GW(e), and is assumed to remain constant 58 GW(e) after 2030. The breeding ratios of FR are 1.10 (in a breeding core) and 1.03 (in a break-even core), and there will be a switchover to a low breeding type core according to Pu balance. A FR deployment scenario is shown in Fig. 3.22. The switchover from LWR to FR will be completed around 2110. With a sodium-cooled reactor cycle, it will take about 60 years to complete the switchover from a LWR to a FR cycle, which is comparable to a typical LWR plant life, and therefore it will enable independence from a uranium resource by the beginning of the 22nd century. Fig. 3.20 shows accumulative natural uranium demand. The accumulative uranium demand in 2200 for a once through scenario and a Pu multi-recycling in a LWR scenario is approximately 2.3 million ton U and 1.9 million ton U, respectively. This corresponds to 16 % and 13 % of total available conventional resources (14.8 million tons U) [10], respectively. In the FR deployment scenario, accumulative uranium demand decreases to around 0.7 million tons U (about 5 % of total conventional resources). The present Japanese uranium demand accounts for 10 – 15 % of the total world demand and its share will decrease to approximately 5 % in the second half of this century because of an increase in nuclear power electricity demand in other mainly Asian countries. 100
Total capacity is about 58 GW e After 2030. 80
60 LWR (UO2 fuel) 40 LWR (MOX fuel) FR (BR. 1.03) 20
Nuclear Capacity (GWe) FR (BR. 1.1) 0 2000 2050 2100 2150 2200 Year
FIG. 3.22. Transition scenario from a LWR cycle to a FR cycle in Japan. Fig. 3.23 illustrates minor actinides (MA) accumulation in high-level radioactive waste which consists of spent fuels for direct disposal and the vitrified glasses for Pu recycle scenarios.
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600 R FBR（Sodium-MOX) W L in 500 LWR(Pu recycling) g in LW R(once-through) cl cy re gh 400 u u P ro e -th nc o 300 R LW
200
FBR cycle system 100
MA MA wastein andspent fuels (tonHM) 0 2000 2050 2100 2150 2200 Year FIG. 3.23. MA accumulation in high-level radioactive waste in Japan.
Direct disposal of LWR spent fuel (both UO2 and MOX) will increase MA accumulation. The MA accumulation in vitrified glasses changes drastically if MA is assumed to be recovered in future LWR reprocessing plants. On the other hand, in the FR deployment scenarios, as the MA recovered from LWR reprocessing plants are recycled in FR fuels, the MA accumulation in vitrified glass is estimated to be kept below 100 tons HM throughout the next century. The nuclear power generation capacity of Japan was supposed to become 58 GW(e) in the future, and a long-term mass flow analysis for representative nuclear scenarios was carried out. From the viewpoint of a reduction of environmental burdens, a significant reduction of actinides (U, Pu, and MA) in high-level radioactive waste is expected in a FR cycle deployment scenario. Most of actinides can be managed within that FR cycle. A FR cycle deployment scenario is superior to any other scenario regarding a reduction of environmental burdens and natural uranium demands. Therefore, such a cycle is considered to contribute to the preservation of the environment and to a sustainable utilization of nuclear power.
3.4.5. National approach identifying fast reactors designs and relevant fuel cycles
Feasibility study on commercialized fast reactor cycle systems Phase I of the Japanese feasibility study (F/S) was carried out during Japanese fiscal year JFY 1999 to 2000, and phase II was launched in April 2001. Based on the results of phase I, representative concepts of fast reactor systems have been screened regarding the use of various coolants, such as sodium, lead-bismuth, helium and water, introduction of oxide, nitride and metal fuels and fuel reprocessing methodologies such as advanced aqueous, oxide electro-winning and metal electro-refining systems. In phase II (JFY 2001−2005), in order to clarify promising fast reactor cycle concepts, investigation of design concepts and development of key technologies needed for quantitative evaluation were carried out for several concepts screened in phase I. The final report of the phase II study was published in March 2006. Based on the comprehensive evaluation in the F/S, the fast reactor cycle system composed of a sodium-cooled fast reactor with a MOX fuel core, an advanced-aqueous reprocessing method and a simplified-pelletizing fuel fabrication method was recommended as the mainline choice for the most promising concept as shown in Fig. 3.24 and 3.25.
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ODS steel cladding tube for high Prevention of sodium burnup Secondary pump chemical reaction • Complete adoption of a double-walled piping system Reduction in material amount and building volume through Steam generator • Steam generator with straight the adoption of innovative double-walled heat transfer tube technologies
• 2-loop arrangement for a Inspection and repair simplified plant system technology under sodium • High-chromium steel structural material for a shortened piping length Integrated intermediate Enhancement of reactor heat exchanger • Integrated intermediate heat with primary pump core safety exchanger with primary • Passive reactor shutdown pump for a simplified primary system and decay heat cooling system removal by natural circulation • Compact reactor vessel Reactor vessel • Recriticality free core concept during severe core damage
FIG. 3.24. Concept of the sodium-cooled reactor in Japan.
Also a R&D plan until around 2015 and the key issues to be solved for commercialization were proposed in the final report. Following publication of the F/S result, the ministry of education, culture, sports, science and Technology (MEXT) assessed the result of the F/S by October 2006. As a result, the abovementioned concept was confirmed as the main concept which should be developed principally toward its commercialization. Key R&D issues necessary to commercialize were defined.
Disassembling and Shearing
In-cell fuel fabrication enabling low decontamination and MA Pin fabrication and recycle U crystallization process assembly of bundle that can dramatically reduce Dissolution the extraction process flow Die lubricating-type molding without lubricant- mixing Single cycle co-extraction of U, Pu and Np with low Molding and Sintering decontamination Crystallization
Powder mixing process is removed by adjusting Pu No Purification processes of U content at solution state and Pu because the recovery in Co-extraction Adjustment of Pu content low-decontamination process is permitted. Denitration, Calcination & Reduction, Granulation
Adjustment of Pu content at MA recovery by using extraction solution state is enabled by chromatography that allows the High-level integrating reprocessing use of compact components and liquid waste and fuel fabrication plant a lower amount of secondary waste MA recovery by extraction chromatography FIG. 3.25.Concept of a combination of an advanced aqueous reprocessing system and a simplified pelletizing fuel fabrication system in Japan.
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Based on the conclusions of the F/S and a check and review by the government, a “fast reactor cycle technology development” (FaCT) project had been launched as an advanced stage toward commercialization of fast reactor cycle technology.
Program of the FaCT project Fig. 3.26 shows an outline of commercialization of fast reactor cycle technology in Japan. In the FaCT project, a design study and related experiments regarding the main concept will be implemented in order to present the conceptual designs of commercial and demonstration fast reactor cycle facilities by Japanese fiscal year JFY2015 with a development plan to realize them. Key R&D issues, 13 issues for FR and 12 issues for the fuel cycle were defined by the F/S and its assessment, and all innovative technologies are planned to be developed by JFY2015. The goal of R&D by JFY2010 is to decide the adoption of innovative technologies by judging their applicability. A program to demonstrate engineering-scaled equipment for a conceptual design will be also investigated. After JFY2015, the FaCT project will step up to the introduction stage of a first system demonstration. The demonstration fast reactor will start to operate in around JFY2025. Before JFY2050 a commercialized fast reactor cycle system will be deployed based on the experience of the demonstration fast reactor cycle system.
FIG. 3.26. Outline of commercialization of fast reactor cycle technology in Japan.
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3.5. REPUBLIC OF KOREA
3.5.1. National planning of capacities and structure of energy sector with determination of the share of nuclear power and fast reactors
Installed capacity and electricity generation There is an ever increasing demand for electricity in the Republic of Korea. Economic growth and the desire for a better quality of life are the main reasons for the increase. Demand has increased by about ten times in the last 25 years since 1980 with an average annual growth rate of 9.6 %. The estimated average annual growth rate is estimated to be 2.5 % during the period from 2006 to 2017. Therefore electricity generation capacity needs to be increased continuously in order to meet demand. However, the availability of conventional energy resources is extremely limited in the Republic of Korea: no oil, little natural gas, and limited sites for hydro power. There are coal deposits, but their utilization is minimal due to harmful environmental impacts. Consequently, domestic natural energy resources had a 2.8 % share of gross electricity generation in 2005. Since nuclear energy is less dependent on natural resources and more technology-intensive, nuclear power plants play a very important role in achieving energy self-reliance and in stabilizing electricity price. Fig. 3.27 and 3.28 show the nuclear share of total installed capacity and electricity generation in the Republic of Korea during the past 9 years. In the year 2005, nuclear power plants represented 28 % of the total installed capacity, but generated as much as 40 % of total electricity.
Electricity Generation Installed Capacity GWe % 40 60 35 40 30 20 25 0 20 97 98 99 00 01 02 03 04 05
Nuclear 10.3 12 13.7 13.7 13.7 15.7 15.7 16.7 17.7 Total 41 43.4 47 48.5 50.9 53.8 56.1 60 62.26 Nuclear Share, % 25.1 27.7 29.2 28.3 27 29.2 28 27.9 28.4
FIG. 3.27. Installed electricity generation capacity in the Republic of Korea.
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Annual Electricity Generation GWh % 400,000 50 300,000 45 200,000 40 100,000 35 0 30 97 98 99 00 01 02 03 04 05
Nuclear 77,100 89,700 103,100 109,000 112,100 119,400 129,700 131,964 146,779 Total 224,400 215,300 239,300 266,400 285,200 306,500 322,300 341,981 364,639 Nuclear Share, % 34.3 41.7 43.1 40.9 39.3 38.9 40.2 38.6 40.3
FIG. 3.28. Annual electricity generation in the Republic of Korea.
Nuclear power plants in the Republic of Korea The role of nuclear power in electricity generation is expected to become more important in the Republic of Korea in the years to come due to an increasing electricity demand and limited natural resources. Since the first commercial nuclear power plant Kori Unit 1 started its operation in 1978, there are now 16 PWRs (6 OPRs) and 4 PHWRs in operation as of May 2006. Fig. 3.29 shows the locations of the nuclear power plant sites in the Republic of Korea. According to the "basic plan for long-term electricity supply and demand” by the ministry of commerce, industry and energy a total of eight new nuclear power plants will be constructed by 2017. The construction of four OPRs will begin before the end of 2006: two at the Kori and two at the Wolsong site. The first concrete of four more APR1400s are scheduled to be poured between 2011 and 2012. With these eight additional plants in the nuclear fleet, nuclear power will represent about 33 % of the total installed power capacity and 47 % of the total electricity generation in 2017.
FIG. 3.29. Current status of nuclear power plants in the Republic of Korea.
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With a continuous expansion of nuclear power capacity, overall PWR spent fuel storage capacity is foreseen to be saturated by 2016, even if considering the expansion of spent fuel storage pools at each nuclear power site. In addition, it is difficult to select a waste disposal site from the viewpoint of public acceptance. In 1998 the Korean Atomic Energy Commission (KAEC) decided to construct one away-from-reactor spent fuel storage facility by 2016. Spent fuels will be stored there until a policy is established for the spent fuels. The radioactive waste disposal is an impending challenge in the Korean nuclear society. Sodium cooled fast reactor (SFR) deployment scenarios coupled with PWRs by TRU recycling from PWR and PHWR spent fuel clearly show that re-use of PWR spent fuel by coupling (mainly) PWRs and SFRs can drastically reduce the amount of domestic PWR spent fuel and minor actinides regarded as high level radioactive waste. As a result, the PWR and SFR coupled closed nuclear fuel cycle can reduce the burden of radioactive waste disposal. In this regard, fast reactors are perceived a promising nuclear energy system that can support domestic waste management in the Republic of Korea. It is expected that SFRs are able to achieve a substantial reduction of PWR spent fuel at the first stage of deployment. R&D on fast reactors and related fuel cycles is now being performed in order to provide a technical basis and support to the decision making process.
3.5.2. Brief history and current status of the fast reactor technology development
It has been widely recognized in the Republic of Korea that the fast reactor system is one of the most promising nuclear options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive waste. In response to this recognition, SFR technology development commenced in 1992 when the KAEC approved a national mid- and long-term nuclear R&D program on SFR with the objective to develop key technologies for fast reactors which can meet the goals of sustainability, safety and economic competitiveness. The development efforts focused on basic research of core neutronics, thermal hydraulics and sodium technology. At present, the efforts have been extended towards the development of the basic key technologies of an advanced fast reactor concept called KALIMER-600 (600 MW(e)). Future nuclear power plants should meet several criteria: effective resource utilization, waste minimization and reduced environmental impacts, economic competitiveness, enhancement of safety and reliability, proliferation resistance, and physical protection. In order to meet these criteria, the ministry of science and technology (MOST) established the comprehensive nuclear energy promotion plan in 2001 which set out the basic framework of nuclear R&D. The plan suggested the development of liquid metal reactor technologies for an efficient utilization of uranium resources with emphasis put on basic key technologies. The work scope of the present “liquid metal reactor design technology development” project (LMRDTD) under this plan include reactor design studies, the development of computational tools, and the development of sodium technologies. According to the nuclear technology roadmap established in 2005, a sodium cooled fast reactor (SFR) is chosen as one of the most promising innovative nuclear reactor systems which are deployable by 2030. The LMRDTD project is being carried out by the Korean Atomic Energy Research Institute (KAERI) as shown in Fig. 3.30.
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Phases 1&2 (1997-2001) Phase 3&4 (2002-2006) Development of Basic Technology Development of Key Technologies / Conceptual Design / Advanced Conceptual Design
Advanced
KALIMER-150 Conceptual Design KALIMER-600 Conceptual Design - Proliferation-Resistant Core Design and Fuel Cycle - Design requirements Conceptual - Improvement of Economics Conceptual - Design Specification Design and Safety Design - Performance Analysis Report - High Temperature Report - Safety Analysis Structural Analysis - Safety Analysis
Development of Development and Key Computer Codes Improvement of / Validation of Computer Codes Computer Codes
Experiments for Basic Experiments Model Validation
FIG. 3.30. Liquid metal reactor technology development project plan in the Republic of Korea. During phase 1 and 2, basic technologies and a conceptual design of the Korean advanced liquid metal reactor KALIMER-150 (150 MW(e)) have been developed. During phase 3 and 4 from the year 2002 to 2006, key technologies and the conceptual design of KALIMER-600 (600 MW(e)) are being developed with emphasis on: • Proliferation-resistant core design and fuel cycle concepts; • System design for enhancement of economics and safety; • High temperature structural analysis technologies; • Safety analysis methodologies. R&D activities have been performed to develop design methodologies, computational tools for design and analysis, advanced concepts including a supercritical CO2 Brayton cycle energy conversion system, and sodium technology with the construction of several sodium loops and test facilities. In addition to the R&D activities for the SFR technology development, design studies of a 900 MW(th) lead-alloy cooled breakeven/transmutation core with nitride and metal fuel have been performed to investigate the feasibility for power plants as well as actinide/fission product transmutation as an alternative fast neutron system to SFR. The conceptual design of KALIMER-600 (Korean advanced liquid metal reactor with 600 MW(e)) will be completed in February 2007 when the project ends.
3.5.3. Scenarios for development of a closed nuclear cycle with fast reactors
Recently, international programs to develop innovative nuclear energy systems, including both reactors and fuel cycle technologies, have been initiated under the names of the Generation-IV International Forum (GIF) started by US DOE and the International Project on
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Innovative Nuclear Reactors and Fuel Cycles (INPRO) organized by the IAEA. The Republic of Korea has been participating in both programs as well as performing its own comprehensive R&D programs. As a result of all these vigorous research activities, future nuclear energy systems in most countries will consist of various mixes of nuclear reactors and fuel cycle technologies, in order to prepare for increasing energy consumption in the future and to contribute to sustainable growth towards a better quality of human beings’ life. The Republic of Korea also continues to develop a future nuclear energy system with a goal to achieve economic competitiveness, reduction of environmental burden through waste minimization, sustainable resource utilization, technical feasibility and nuclear non- proliferation. To meet these goals, several different reactor types, such as a sodium cooled fast reactor (SFR), a very high temperature gas cooled reactor (VHTR) and an accelerator driven subcritical system (ADS), are taken into consideration and various nuclear fuel cycles are studied in the Republic of Korea. As one of them, the design concept of KALIMER has been developed under the national medium- and long-term nuclear R&D program. KALIMER is a pool type, sodium cooled fast reactor (SFR). It is loaded with metal fuel recycled from PWR spent fuel. In connection with the nuclear fuel cycle, the DUPIC (direct use of spent PWR Fuel in CANDU reactors) scenario involving partial recycling of PWR spent fuel materials in CANDU PHWR reactors is also investigated in the Republic of Korea. However, in the following one investigates the combination of PWR and KALIMER mainly for the INPRO Joint Study on CNFC‒FR.
Adopting of scenarios and their evaluation As stated before reprocessed PWR spent fuel is used in KALIMER. KALIMER spent fuel is also reprocessed and continuously recycled. In this study, three kinds of nuclear scenarios are adopted and the total amount of accumulated spent fuel and minor actinides are evaluated for a period from 2001 to 2100: • Case-1: Once-through strategy; The spent fuel from PWR and CANDU is directly disposed. • Case-2: Partial deployment of KALIMER; PWR spent fuel reprocessing begins in 2020. PWR will be gradually replaced with KALIMER from 2030. Then KALIMER will reach a 21 % fraction of total nuclear energy capacity in 2100. • Case-3: Full deployment of KALIMER; PWR spent fuel reprocessing begins in 2020. KALIMER is deployed gradually instead of PWR from 2030 on and will reach a 76 % fraction of total nuclear energy capacity in 2100. After 2090 to meet demand all new constructed plants will be KALIMER. In 2005, 16 PWRs and 4 PHWRs are in operation in the Republic of Korea. The electricity generation capacity by nuclear energy was 13 716 MW(e) in 2000, 17 716 MW(e) in 2005 and will become 25 237 MW(e) in 2015 according to the Korean long term electricity supply and demand plan. The expected yearly growth rate of nuclear capacity between 2000 and 2015 years is about 4.3 %. The nuclear share will be about 33.0 % of the total generation capacity of electricity in 2017. When one assumes a yearly growth rate of 1.0 % after 2015, the nuclear power capacity (demand) is expected to increase to 58.8 GW(e) by 2100. Fig. 3.31 shows the expected growth of nuclear electricity demand in the Republic of Korea.
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70 60 50 40 (GWe) 30 Nuclear Energy
Demand 20 10 0 2000 2020 2040 2060 2080 2100 Time(yr)
FIG. 3.31. Predicted growth of nuclear electric generation in the Republic of Korea. Table 3.7 shows the typical data of PWR and CANDU, and assumptions of KALIMER data used as input for code analysis. TABLE 3.7. ASSUMPTIONS USED IN ANALYSIS Items PWR CANDU KALIMER Power (GW(e)) 1.00 0.7 0.6 Load factor 0.85 0.85 0.85 Burnup (GW(th)d/t) 40 7 93.9 Thermal efficiency 0.35 0.35 0.393 Fuel cycle length (years) 1.5 0.7 1.67 Number of batches 3 1 4 Fresh fuel fraction - Pu 0 0 0.1981 - MA 0 0 0.0066 - U 1.0 1.0 0.7945 Spent fuel fraction - Pu 0.01201 0.01201 0.1728 - MA 0.00125 0. 00149 0.0063 -U 0.93500 0. 93500 0.7286
Reprocessing Method Pyro Pyro Pyro Capacity(kt(HM)/a) 0.5 0.1 1.0 Ex-core time (years) - Construction 4 4 4 - Enrichment 1 0 0 - Fabrication 1 1 2 - Cooling 5 5 5 - Reprocessing 1 1 1 Reactor Life (years) 40 40 60 Others Typical CE type Operation only until TRU burner PWR 2045
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The DYMOND code was used in this analysis. This code employs the ITHINK dynamic modeling platform to create 100-year dynamic evolution scenarios for postulated nuclear energy systems based on growth of nuclear energy demand. Reactor life time is assumed to be 40 years for both PWRs and CANDUs, and life time of KALIMER is assumed to be 60 years. In this scenario, all CANDU PHWRs are assumed to be shut down after their designed life time and no more CANDU would be constructed thereafter.
Results of DYMOND analysis Based on the results of scenarios’ evaluation with DYMOND, material inventories are summarized in Table 3.8. TABLE 3.8. INVENTORIES OF MATERIALS IN NUCLEAR FUEL CYCLE (kilo tons; kt) IN THE SELECTED THREE SCENARIOS
Total SF U in SF Total Pu Total MA Total FP Case Year (kt) (kt) (kt) (kt) (kt)
2035 29.7 27.7 0.36 0.04 1.6 Case-1 2050 44.1 41.2 0.53 0.06 2.3 2070 61.6 57.6 0.74 0.08 3.2 2100 94.0 87.9 1.1 0.12 4.9 2035 13.4 12.5 0.15 0.04 1.6 Case-2 2050 11.5 10.7 0.30 0.06 2.3 2070 6.4 6.0 0.46 0.08 3.2 2100 8.7 8.1 0.51 0.10 4.8 2035 13.4 12.5 0.15 0.04 1.6 Case-3 2050 11.4 10.7 0.24 0.05 2.9 2070 5.0 4.6 0.01 0.0 9.7 2100 0.0 0.0 0.0 0.0 42.5 SF=Spent fuel; MA=Minor actinides; FP=fission products.
Fig. 3.32 shows the spent fuel accumulation when a once-through strategy and no- reprocessing is adopted (case 1).
100 K PWR SF 80 CANDU SF Total SF 60
40
20 Amount of Spent Fuel ( 0 2000 2020 2040 2060 2080 2100 Time(yr)
FIG. 3.32. Accumulation of spent fuel (SF) in a once-through strategy (case 1).
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Fig. 3.33 displays the fraction of each type of power plant within total nuclear energy capacity development assumed for case 2 and 3. It is assumed that no new CANDU reactors will be constructed and all of them be retired in 2045.
100
80
CANDU 60 PWR Case-2 K-600 Cas e-2 40 PWR Case-3
Capacity (%) in (%) Capacity K-600 Cas e-3
Total Nuclear Energy . 20
0 2000 2020 2040 2060 2080 2100 Time(yr)
FIG. 3.33. Capacity fraction of each type of power plant (case 2 and 3). The amount of spent fuel reduction for case-2 and case-3 can be found in Fig. 3.34 and Fig. 3.35, respectively.
100 PWR SF (Kt) 80 CANDU SF Total SF before 60
40 of Spent Fuel Fuel Spent of
20 Amount 0 2000 2020 2040 2060 2080 2100 Time(yr)
FIG. 3.34. The amount of spent fuel reduction for case-2.
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100 PWR SF (Kt) 80 CANDU SF Total SF before 60
of Spent Fuel Fuel Spent of 40
20 Amount 0 2000 2020 2040 2060 2080 2100 Time(yr)
FIG. 3.35. The amount of spent fuel reduction for Case-3.
Fig. 3.36 shows the amount of minor actinides (MA) disposed as high level waste.
0.20
MA Case-2 0.16 MA Case-3
(kt) 0.12 before-M A( CAs e-1)
0.08 Amount
0.04
0.00 2000 2020 2040 2060 2080 2100 Time(yr)
FIG. 3.36. The amount of minor actinides (MA) disposed as high level waste (Case-1, -2, -3).
Conclusion of the scenario evaluation The evaluation of the PWR/KALIMER coupling scenario shows that the amount of domestic spent fuel from PWR and CANDU can be drastically reduced when KALIMER is deployed from 2020 till 2100. The amount of PWR spent fuel will not be increased any more beyond 2020. And the amount of minor actinides (MA) disposed as high level waste is decreased to zero, if KALIMER is replacing PWR completely. The results of the PWR/KALIMER coupling study will be used as basic input for a study of the management of domestic PWR spent fuel, a fuel cycle study in pursuit of reducing spent fuel storage and high level waste.
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3.5.4. National approach to identifying fast reactors designs and relevant fuel cycles
The third national mid- and long-term nuclear R&D program is currently being finalized to establish a detailed R&D plan for the years from 2007 to 2016. The objective of the third SFR R&D program is to develop a conceptual design of a SFR consistent with the Generation-IV SFR system research plan. It is envisioned that SFR technology development in the Republic of Korea will enter a new phase from 2007 by participating in the Generation-IV SFR collaboration. The R&D endpoints, which are specified in the Generation-IV technology roadmap report, will be utilized to define design activities and deliverables. The KALIMER-600 design will serve as a starting point for this effort. A new advanced design will be developed to include features with a potential to better meet the Generation-IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. Based upon experiences gained during the development of the conceptual design for KALIMER, KAERI is planning to develop advanced design concepts. There are three main categories of R&D activities under consideration: 1) Conceptual design of an advanced SFR; 2) Development of advanced SFR technologies necessary for its commercialization, and 3) Development of key technologies. For the development of advanced technologies, there are four areas: safety, fuels and materials, reactor systems and balance-of-plant. In addition, R&D will be performed for the improvement of economics and assurance of safety. For the development of the key technologies, the main focus will be put on validating computational tools and developing sodium technologies. Timely, inexpensive, safe, proliferation resistant, and environmentally friendly measures should be introduced for the management of spent nuclear fuel. The closed nuclear fuel cycle system with SFRs based on pyroprocessing technology is expected to meet these challenges. R&D activities on relevant fuel cycles will be continued to provide a technical basis and support to the decision making process for a domestic plan of spent fuel management.
3.6. THE RUSSIAN FEDERATION
3.6.1. Brief history and current status of fast reactor technology development
Basic option: validated and implemented sodium cooled fast reactor technology The Russian Federation possesses unique R&D potential for fast reactors, including long-term experience of successful operation of BN-type reactors: experimental reactors BR-2, BR-5/10 and BOR-60; demonstration plant equipped with a BN-350 reactor; and a commercial plant with a BN-600 reactor (Fig. 3.37).
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FIG. 3.37. The reactor BN-600 at the Beloyarsk NPP site in the Russian Federation. The sequence of sodium cooled fast reactor (SFR) construction and carrying out R&D work for their justification in the Russian Federation (USSR) is as follows: BR-1 (1954) → BR-2 (1956) → BR-5 (1959) → BR-10 (1973) → BOR-60 (1969) → BN‒ 350 (1973) → BN-600 (1980) → BN-800 (under construction) → BN-1600 (conceptual design) → BN-1800 (design proposal). As a result, by now the largest experience in SFR operation in the world has been accumulated in the Russian Federation (125 reactor-years, i.e. 40 % of the world total SFR reactor-years) [13]. The extent of mastering of any nuclear technology can be characterized by the following parameters: • Operating time of components by now; • Load factor; • Radiation impact on the environment; • Experience gained in control of accidents typical for this nuclear technology; • Availability of repair technologies. As shown in Table 3.9, time of components’ operation without overhaul is as high as 300 thousand hours. TABLE 3.9. ACHIEVED OPERATION TIME OF SFR COMPONENTS WITHOUT OVERHAUL (HOURS) Reactor BR-10 BOR-60 BN-350 BN-600 components (1959) (1969) (1973) (1980) (Start-up year) Stationary components (hours) Reactor Vessel 150 000 200 000 170 000 130 000 Primary Pipelines 300 000 200 000 170 000 130 000 Sodium Pumps 170 000 130 000 100 000 100 000 (hours) (electromagnetic) (centrifugal) (centrifugal) (centrifugal) Intermediate Heat 300 000 200 000 170 000 130 000 Exchangers (hours) Steam Generators - various design 150 000 105 000 (hours) pilot SG
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BN-800 reactor [14] is the most advanced Russian design fast sodium-cooled reactor – design of this nuclear power unit has been licensed and its construction is underway as a fourth unit at Beloyarsk NPP site. Since 2005 construction of BN-800 reactor is funded completely at the expense of the Russian federal budget. It is expected by ROSATOM that the BN-800 will be commissioned in 2012 and will start its operation using MOX fuel from the very beginning. BN-800 reactor design (Fig. 3.38) inherits all the main principles implemented in the design of its prototype – the operating BN-600 reactor.
1. Vessel 2. Guard vessel 3. Reactor core 4. Pressure chamber 5. Confinement device 6. Vault 7. Main circulation pump 8. Upper fixed shielding 9. Large rotating plug 10. Central rotating plug 11. Protection hood Figure 3.37. BN-800 reactor design 12. Reloading mechanism 13. Small rotating plug 14. Intermediate heat exchanger
FIG. 3.38. BN-800 reactor design.
The most important difference between BN-600 and BN-800, besides an increased installed capacity, is the capability of routine reactor operation with MOX fuel and a design concept of using a closed nuclear fuel cycle. The BN-800 fast reactor has been significantly improved with regard to safety due to a number of design and engineering solutions, as compared to the operating BN-600 reactor. Major characteristics of alternative core designs of BN-800 are summarized in Table 3.10.
TABLE 3.10. MAIN CHARACTERISTICS OF CORE DESIGN OPTIONS FOR A BN-800 REACTOR
Option I Option II Option III Parameter Initial Unified by fuel Nitride design with BN-600 Core height, cm 88 90 88 Number of FA 565 584 702 Fuel element diameter / 6.6/0.4 6.9/0.4 7.2/0.45 Wall thickness, mm Volume fraction of fuel, % 38.8 42.9 45.7 Fuel type (U+Pu)O2 (U+Pu)O2 (U+Pu)N Effective density, g·cm-3 8.6 8.6 12.3
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Critical loading, kg Pu 2710 2906 4430 Fuel loading, t 14.3 16.8 30.0 Burnup, % - maximum 10 10 8,3 - average 6.5 6.6 5.4 Core life, eff. days 3*140 3(4)*146 5*140 Blanket: - radial UO2 steel steel - bottom axial UO2 UO2 steel Reactor breeding ratio 1.0 0.957 1.045 Core breeding ratio 0.73 0.813 1.045 Burnup margin, % 3.4 2.7 0.1 Annual plutonium 150 50 0 consumption, kg
Perspective designs of sodium cooled fast reactors BN-1800 fast reactor [15] Analysis of design and experience of operation of Russian and foreign reactors (BR-10, BOR- 60, BN-350, BN-600, PHENIX and SUPERPHENIX) and design of the BN-800, BN-1600 and EFR reactors has shown that there are considerable reserves in SFR technology for improvement of technical and economical characteristics and safety. The main trends of improvement of technology are as follows: • Improvement of NPP thermodynamic characteristics (sodium temperature increase and transition to supercritical steam pressure); • Increase of the fuel burn-up (use of materials with improved radiation characteristics); • Efficient use of inherent safety features in combination with engineered passive safety systems; • Increase of integration degree of circuits and components (reactor and steam generator); • Simplification of reactor refueling system and auxiliary systems; • Increase of power unit lifetime. Taking into account the above propositions, design proposals were developed by MINATOM of the Russian Federation for an advanced large size sodium cooled fast reactor – BN-1800 (Fig. 3.39).
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FIG. 3.39. Advanced sodium cooled fast reactor BN-1800.
The basic technical design characteristics of the proposed BN-1800 unit are presented in Table 3.11.
TABLE 3.11. BASIC TECHNICAL CHARACTERISTICS OF A NUCLEAR POWER PLANT WITH A BN-1800 REACTOR
Parameter Value Electric power of reactor unit, MW(e) 1780 Reactor thermal power, MW(th) 4000 Primary sodium temperature, outlet/inlet, 0С 575/410 Secondary sodium temperature, outlet/inlet, 0С 540/370 Third circuit temperature, outlet/inlet, 0С 525/270 Steam pressure at SG outlet, MPa 25.0 Reactor vessel diameter, m 17 Fuel inventory (UPuN), ton 86 Plutonium loading, ton 12 Number of the primary pumps 3 Number of IHX 6 Number of the secondary loops 6 Number of evaporators-super-heaters 6 Evaporator-super-heater power, MW 536 Number of re-heaters 6 Re-heater power, MW 131 Specific metal content in the reactor plant, ton/MW(e 4.4) Basic design of the BN-1600 reactor was used implying significant modifications of flow diagram, parameters and design approaches of the main components. The main results of these modifications are as follows: • Owing to an increase of sodium temperature and use of supercritical steam parameters, it became possible to considerably improve thermodynamic parameters of the steam-cycle and assure gross and net efficiency values equal to 46 % and 43 %, respectively;
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• Application of a completely integrated primary circuit design implying location of all primary components within the reactor vessel, thus completely eliminating radioactive sodium leaks, use of passive safety systems and further increase of efficiency of inherent safety features (including advantages of nitride fuel) have made it possible to significantly increase reactor safety; • Use of a high density nitride fuel with high burn-up has assured a long refueling interval (1.5−2 years) and a load factor of ~0.9 taking into account the increase of components reliability on the basis of experience gained with the BN-600 reactor. • Use of entirely new design approaches to the reactor refueling system, steam generator, auxiliary systems, etc., increase of the fuel burn-up and load factor in combination with high efficiency have resulted in the improvement of technical and economical characteristics of the unit.
Table 3.12 gives data showing the effect of the above measures on the improvement of technical and economical characteristics [15].
TABLE 3.12. COMPARISON OF SPECIFIC CHARACTERISTICS OF DIFFERENT RUSSIAN BN-TYPE SFR BN-600 BN-800 (under BN-1800 Parameters (in operation at construction on (conceptual BNPP) BNPP site) study) Power of the unit, MW • thermal 1470 2100 4000 • electric 600 880 1780 Technical and economical characteristics (with respect to the BN-600 reactor) • relative specific metal content (t/MW(e)) in reactor plant 1 0.7 0.33 • relative specific capital cost for NPP (single unit NPP) 1 0.9 0.48 Reactor BN-GT-300/100 [16] Another concept of fast reactor technology development is a two-circuit module-type transportable nuclear power plant with a small and medium capacity (up to 300 MW(e)) and with a fast-neutron sodium-cooled reactor and gas-turbine converter ‒ BN-GT-300 (Fig. 3.40).
FIG. 3.40. Nuclear power plant with a BNGT-300/100 reactor.
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Perspective designs of fast reactors with heavy liquid metal coolant A change-over to a two-circuit layout and a gas-turbine cycle may provide, on the one hand, complete implementation of sodium coolant benefits for cooling a reactor core with a fast- neutron spectrum at enough high temperature level; on the other hand, it will minimize sodium coolant drawbacks due to its chemical activity with air and water (i.e. primary circuit should be completely integrated; heat removal system should operate without water inventory). Analysis of this concept of a nuclear power plant demonstrated preliminary essential cost- efficiency benefits as compared to known power plant types. A two-circuit facility unifies both the idea of a fast-neutron sodium-cooled reactor and a medium-temperature close-circuit gas turbine facility. Possible benefits of this approach could be: simplification of power unit design; reduction of specific consumption of materials and specific costs of the power unit construction; reduction of production and transportation time for major plant components; reduction of construction time and assemblage activity at destination site; and a number of other issues related to economy, safety, and proliferation resistance. During several past decades ROSATOM organizations conducted R&D on fast reactors with heavy liquid metal coolants. The developed advanced reactor designs include the lead- bismuth cooled SVBR-75/100 reactor (Fig. 3.41) being developed by SSC RF-IPPE and EDB GIDROPRESS [17], and the lead cooled BREST type reactor (Fig. 3.42) developed by NIKIET [18].
FIG. 3.41. Modular type SVBR 75/100 fast reactor and possibilities of its multi-purpose application. Use of heavy liquid metal coolants (HLMC) based on lead in a fast reactor allows obtaining very important features of reactor self-protection with respect to a number of severe accident types; this ability is stipulated by natural properties of HLMC, i.e. a high boiling temperature and chemical inertness regarding interaction with water and air. By a proper reactor design certain types of severe accidents could be practically eliminated.
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FIG. 3.42. BREST OD-300 (left) and BREST-1200 (right) lead-cooled fast reactors. General remarks on design of a closed nuclear fuel cycle with fast reactors A closed nuclear fuel cycle of fast reactors with plutonium-based fuel has a principal difference compared to an open nuclear fuel cycle of thermal neutron reactors. The difference is as follows: actual development of nuclear power occurs with absolute predominance of thermal reactors and open fuel cycles; this situation will most probably continue as a minimum until the middle of the 21st century. However, in the future (after the century) a system with thermal reactors and an open fuel cycle may face a shortage of fuel supply due to limited natural uranium sources. However, a problem of the fuel supply for systems with fast- neutron reactors and a closed fuel cycle should not arise, because the supply problem could be solved practically in full extent by using “major production waste”, i.e. spent nuclear fuel of thermal-neutron reactors. The physics of fast reactors is characterized by the fact that fast reactors are capable of producing new nuclear fuel while using both surplus plutonium and minor actinides generated in the thermal reactors and depleted uranium tails from the enrichment stage of the thermal reactor fuel production.
Closed nuclear fuel cycle for mainline direction of fast BN type reactors Presently, the required commercial infrastructure for implementing a MOX fuel production and spent fuel reprocessing is absent in the Russia Federation; nevertheless, related studies are being conducted in the frameworks of extensive technical, technological, and experimental R&D programs; they resulted in demonstration pilot facilities which simulate all operation parameters of a large-scale production. The activity takes into account experience of Western Europe where MOX fuel production and reprocessing has reached commercial scale. The capability was demonstrated by calculations to operate a fast reactor with MOX fuel based on both reactor grade plutonium and weapons-grade plutonium in a wide range of plutonium compositions without any major influence on reactor characteristics. At the BN-600 reactor a program of experiments has been carried out to confirm the feasibility of using pellet mixed uranium-plutonium (MOX) fuel in BN-600 and BN-800 reactors; test runs for vibro- compacted MOX fuel are under completion. A starting point for comparing different fuel cycle modifications is a fuel production concept, which is nowadays implemented in several countries and established as a base for the developed project of the BN-800 reactor. Essential elements of the concept are a pellet technology for MOX fuel fabrication and aqueous radiochemical reprocessing of spent fuel
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with partitioning uranium and plutonium (PUREX process) (Fig. 3.42). A closed fuel cycle technology as a basis for comparison and review could be applied for advanced BN-type reactors (BN-1800, BNGT). Advanced nuclear fuel cycle for BN reactors with MOX fuel Within the framework of an advanced nuclear fuel cycle for BN-type reactors, MOX fueled reactor cores (in prospect – nitride fueled reactor cores) with a breeding ratio slightly above 1 are under investigation to ensure a safe reactor operation in a closed fuel cycle without any external plutonium feed. In this case: • Important institutional/technical support would be achieved for a nuclear material nonproliferation regime; • Low-level waste processing of spent nuclear fuel is implemented and radwaste removal becomes radiation-balanced.
Fig. 3.43 presents a flowchart of a closed nuclear fuel cycle for the BN-800 reactor.
Fresh fuel storage Irradiated FSA BN-800 reactor at NPP site cooling pond
Fuel, fuel elements, Irradiated FSA Fresh FSA Transportation of and FSA radiochemical transportation irradiated FSA fabrication plant reprocessing plant
FIG. 3.43. Flowchart of the closed nuclear fuel cycle for a BN-800 reactor.
Design of reactor cores: Nowadays, advanced designs of the reactor cores of the reactor BN- 800 are under development. It is assumed that first oxide fuel will be applied and nitride fuel is a prospective version. Both reactor cores have standardized dimensions of the fuel rod and fuel subassembly (FSA), the same number of FSA, and the same number and location of control and protection system rods. FSA of a MOX fueled reactor core has a lower axial blanket made of depleted UO2. FSA of a nitride-fueled reactor core has a steel lower axial blanket. Two enrichment-grade fuels (for a low enrichment core zone, LEZ and a high enrichment core zone, HEZ) are implemented to flatten the energy release profile. Both reactor cores have a breeding ratio slightly above 1, thus enabling a reactor operation in a closed fuel cycle without any external plutonium feed. Spent fuel storage pool: The pool is a reservoir with steel-made walls filled with water. Spent FSA are emplaced into the pool after unloading from a tank of spent FSA. The spent fuel pool is located inside in the reactor containment building. Spent FSA transportation: After cooling down spent FSA are loaded into a shipping-package set. A shipping-package set represents a cylindrical reservoir with thick steel walls (300−350 mm) for radiation protection. After emplacing the FSAs housing into the container, the shipping-package set is closed by a cap and transported in a special railway wagon to a reprocessing plant. Radiochemical plant for reprocessing spent FSAs: Spent reactor fuel is supposed to be reprocessed at the RT-1 plant site; this plant is located at “Mayak” as far as 250 km from the
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Beloyarskaya nuclear power plant and possesses the infrastructure needed, including radwaste processing plant for wastes having different levels of activity. The plant structure is designed for hydro-metallurgic technology and use of pelletized fuel. The VVER reactors’ spent nuclear fuel reprocessing could be carried out by 3 process lines at the RT-1 plant; one of them could be applied to reprocess fuel of the reactor BN-800 after a necessary modernization. Recently, the process flowchart of the RT-1 plant is being updated. Therefore there is an opportunity to include necessary modifications into the updating schedule to enable reprocessing of spent nuclear fuel of the BN-800 reactor. Spent FSA with MOX fuel will be reprocessed together with the lower axial blanket. The purpose of reprocessing is to extract specific fission products from spent fuel materials (as high as ∼0.8 % of minor actinides remains in the fuel). These modifications have been accomplished during development of process equipment and operation related with chemical reprocessing of spent nuclear fuel, fresh fuel fabrication, as well as transportation of fresh and spent fuels. The plant for fabrication of fuel, fuel rods, and FSA: It is planned that a plant for MOX fuel rods and FSA fabrication (“Complex-300” and its analogue) will be sited at the “Mayak” as well. Fresh FSA transportation: Fresh MOX FSA will be transported from the fuel fabrication plant to NPP site in special shipping-package sets for fresh fuel transportation. A shipping- package set for fresh fuel transportation is a steel-made cylindrical housing which is loaded with one fresh MOX FSA. Fresh fuel storage facility: A storage facility for fresh fuel will be constructed at the site of the Beloyarskaya NPP’s power unit 4; it is a separate building located as far as 100−200 m away from the BN-800 reactor building. On-site transportation of fresh MOX fuel at the Beloyarskaya NPP’s power unit 4 site is performed using special on-site containers. A fuel cycle similar to that considered for the reactor BN-800 could be implemented for the reactor BN-1800 as well. Because of some features of the BN-1800, physical issues of designing advanced reactor cores with mixed oxide and nitride fuel and a breeding ratio slightly above 1 can be solved simpler for the large-scale BN-1800 reactor compared to a BN- 800 reactor. On-site nuclear fuel cycle for BN-800 reactor Analysis of possible spent fuel flows in a nuclear power system involving contemporary and advanced reactors allows the conclusion that in order to optimize fuel cycles, a systematic approach considering all problems need to be taken, i.e., nuclear reactors and fuel cycle facilities should be assessed as a whole system. An on-site closed fuel cycle of the BN-800 reactor is a prospective approach that allows arranging such kind of a system. The uniqueness of this approach is that this kind of a technological solution could be easily introduced into the existing fuel cycle schemes of different power reactor types. Fig. 3.44 demonstrates one of the possible ways of combining nuclear material flows in the frame of a general fuel cycle of the BN-800 reactor. The outlined scheme of a design of a combined nuclear fuel cycle for power reactors, taking into account spent fuel already accumulated in the Russian Federation (from RBMK and VVER reactors) and possibly foreign spent fuel (primarily from PWR reactors), has been considered based on the following technological options: • Gaseous-fluoride reprocessing technology for aged spent fuel (PWR and VVER); • Non-aqueous spent fuel reprocessing methods for spent fuel of RBMK type reactors (compaction/ metallization) to reduce storage area and long-term storage costs.
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COMPLEX FOR PREPARING RADIOACTIVE WASTE FOR CONTROLLED STORAGE
Reactor BN-800 ON-SITE COMPLEX FOR REPROCESSING SNF AND MANUFACTURING FA WITH VIBRO- COMPACTED FUEL RODS
FRESH FUEL ON-SITE STORAGE
FIG. 3.44. Elementary scheme of an on-site closed fuel cycle of a BN-800 reactor based on non-aqueous methods of reprocessing spent fuel and manufacturing of fuel rods by a vibro- compaction method. Advanced closed fuel cycle based on non-aqueous SNF reprocessing methods and manufacturing fuel rods by means of vibro-pack technology Acquired expertise by NIIAR in implementing a closed nuclear fuel cycle for fast-neutron reactors allows considering an alternative innovative closed nuclear fuel cycle based on non- aqueous technologies of spent fuel reprocessing and manufacturing fuel rods by means of vibro-compaction method. Reactor cores: Comments and characteristics outlined for an innovative cycle based on aqueous methods of spent fuel reprocessing procedures are completely valid for that scheme of the on-site closed fuel cycle. The issue is supported by an assumed application of vibro- compacted MOX fuel whose test runs in the reactor BN-600 are being completed. Spent fuel storage pool: It is assumed that a spent fuel storage pool will be operated as an intermediate storage facility before shipping the irradiated FSA for fuel reprocessing; storage time will meet time requirements needed to implement a continuous technological process “reprocessing - fabrication“. Additionally, there is a possibility investigated to eliminate cleaning of spent fuel assemblies from sodium. Transportation of spent fuel subassemblies: Arrangement of an on-site fuel management complex at the nuclear power plant site results in a drastic simplification of spent fuel assemblies’ transportation to the reprocessing facility. Radiochemical plant for reprocessing spent fuel subassemblies, fuel, fuel rod, FSA manufacturing plant: A complex innovative solution (on-site plant for closed fuel cycle) is considered. The plant design involves engineering solutions as follows: • Spent fuel digestion methods (mechanical, thermal dismantling of FSA); • Pyrochemical methods of spent nuclear fuel reprocessing; • Preparation of cathode precipitates and manufacturing of granulated fuel; • Manufacturing of fuel rods using a vibro-compaction method; • Manufacturing of fuel subassemblies; • Preparing of generated waste for controlled storage. The pyrochemical reprocessing flowchart consists of the following principal stages: • Dissolution of initial products or spent nuclear fuel in molten salts; • Recovery of crystal plutonium dioxide or electrolytic uranium/plutonium dioxides from molten salts; • Treatment of cathode precipitates and manufacturing of granulated fuel.
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Operations of the cathode precipitates treatment (trapped salt washout, drying, milling, and classification) do not change fuel’s chemical composition and do not result in any plutonium redistribution or isolation. Impurity content of the finished product meets requirements of the related specifications. Acquired knowledge on molten salt systems and practical expertise on fuel production and processing allows searching for an optimum regeneration flowchart to be applied for an individual type of processed fuel. Fuel rods and FSA: Manufacturing fuel rods with granulated fuel is carried out by means of a vibro-compaction method according to a conventional procedure (in glove boxes or protective chambers); the method has been operated by NIIAR for more than 20 years. Major benefits of the vibro-compaction technology for manufacturing fuel rods are as follows: • Engineering process simplicity and reliability due to a lesser number of technological and control operations thus facilitating automation and remote control; • Possibility to apply any granulate form, both a homogeneous composition and a mechanical mixture; • Reduced mechanical impact of vibro-compacted fuel on the cladding (as compared with a pelletized fuel core); • Relaxed requirements for inner diameter of fuel rod cladding. Creation of the automated remote-controlled processes of manufacturing fuel rods and FSA allow implementing the following issues: • Changeover to new short-term efficient processes of a fuel cycle; • Solving problems of complex production regarding safety and nonproliferation of fissile materials; • Involvement of plutonium (including reactor grade one) into the nuclear fuel cycle. Radioactive waste: Pyrochemical regeneration processes of spent nuclear fuel do not generate liquid radioactive waste; some specific types of solid radioactive waste are generated and treated according to standard methods: compaction, vitrification, and combustion. Safety systems: Development of safety systems is performed in accordance with requirements identified by norms and regulations being in force to provide radiation, nuclear, chemical, fire and physical safety and security. Radiation safety of pyrochemical processes: Radiation safety of pyrochemical processes is ensured through implementation of design solutions based on sanitary norms and rules followed during the design of the radiation-dangerous facilities. Experience and knowledge obtained in course of developing and mastering the reprocessing and granulation processes of mixed fuel allow creating an efficient radiation safety system. A salt system containing dissolved fuel is an ionic liquid. Due to cation-anion interaction all components of the melt have complex forms, therefore the partial vapor pressure of fission product compositions is 102−103 times lower than that of pure fission products. Consequently, transformation of fission products into gaseous phase is limited. Fission products transformed into a gaseous phase have a restricted distribution area and being sublimates of salt chlorides can be effectively removed using specialized filtration devices. Efficiency of the gaseous release purification system is based on the following features:
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• Processing is performed in sealed devices; • Gaseous phase undergoes three stages of purification: immediately at the outlet of a device, at the outlet of a protective chamber, and in the building collector. The chosen technological scheme does not envision direct generation of liquid radwaste, though they result from decontamination of protective chambers, equipment components and other pieces subject to contact with radioactive materials. One of such liquid radwaste sources is the system for surplus chlorine management that envisages cryogen purification, accumulation and reuse of chlorine in the process. Reduction of the contamination level of liquid waste is to be achieved by applying dry treatment of surfaces, evaporation of liquid waste containing considerable quantities of radioactive and fissile materials, and reuse of the dry product in the technological process. Nuclear safety: Nuclear safety problems for the pyrochemical reprocessing technology are addressed in a relatively simple way because the technological scheme does not involve neutron moderators. The processes of salt treatment are developed so that crucible dimensions allow reprocessing of about 50 kg of fast reactor MOX fuel at one time. This limitation is stipulated by the nuclear safety requirements. In case when irradiated materials from both core and axial blankets are reprocessed simultaneously, batch size can be increased without giving raise to criticality issues provided that overall equivalent enrichment of the materials being processed is decreased. It should be also taken into account that presence of chlorine and cesium nuclides in the reprocessed materials will result in a criticality “suppression” effect allowing additional increase of reprocessing facility batch size. Chemical safety: Chemical safety of pyrochemical processes is ensured through the use of gaseous chlorine and application of its designed recycling scheme. Fire safety: The processes do not involve organic materials as extracting substances and diluents, as well as water solutions that regenerate hydrogen. According to the design, the most dangerous process compartments are equipped with special cables with inorganic isolation. Nuclear material control, accounting and protection (MCA&P): Production based on pyrochemical reprocessing technology can be featured by a MCA&P system that meets contemporary requirements of the IAEA. The MCA&P system includes a unified information network that integrates all the technological sections and facilities dealing with processing and accumulation of fissile materials as well as an analytical laboratory and a central process control system. The system ensures continuous control of fissile materials through a number of measures, including: • Weighting of fuel materials after each processing stage; • Analytical determination of actinide content at the main process stages – fuel dissolution in molten salt and fuel certification after fabrication of granulate; • Integral measurement of fissile material content by neutron registration technique when passing from one process stage to another. The concept of co-sited closed fuel cycle facility for BN-800 reactor The co-sited facility is based on the following technologies: • Pyrochemical reprocessing of irradiated nuclear fuel followed by production of granulated fuel material; • Fabrication of fuel elements based on reprocessed granulated fuel materials and using vibro-packing technique;
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• Automated remotely controlled processes of spent fuel reprocessing, granulation, fabrication and control and inspection of fuel elements and fuel subassemblies. Additionally, at the facility the following activities could be performed: • Reprocessing of spent fast reactor fuel and blanket subassemblies; • Manufacturing of lead test batches of fast reactor fuel elements and subassemblies based on “fresh” fuel materials (uranium and plutonium from a PUREX-process or weapon- grade materials) and recycled fast reactor fuel; • Application of up-to-date waste treatment technologies for radioactive waste management including vitrification or solidification; • Recycling of transuranics (Np, Am, and Cm). Source materials for the process are: • Irradiated subassemblies from core, axial and radial blankets of a BN-800 reactor (both uranium oxide and MOX fuel); • In course of facility commissioning, equipment adjustment and tuning can be accomplished using pure oxides of uranium and plutonium. Finished products manufactured at the facility are fuel elements and core fuel subassemblies for a BN-800 reactor (including axial blankets), produced using vibro-packed MOX and UO2 fuel (it is also envisioned that UO2 elements for axial blankets will contain some quantity of Pu and Np). The facility is capable of producing other types of fuels and fuel subassemblies for nuclear reactors: • Mixed uranium-plutonium oxide fuel of different isotopic composition; • Fuel and materials with a high content of minor actinides and plutonium for burning. Influence of purification efficiency level of spent MOX fuel to physical characteristics of a BN-800 type reactor. Tables 3.13 to 3.15 show principal characteristics of granulated fuel and parameters illustrating the efficiency of spent MOX fuel purification from minor actinides and fission products that can be achieved in pyrochemical reprocessing at NIIAR facilities. Main physical characteristics of BN-800 type reactor being important from the viewpoint of criticality and safety include sodium void reactivity effect, Doppler effect, and efficiency of control rods. These characteristics practically do not change when MOX fuel made of “pure” plutonium is substituted with the MOX fuel manufactured based on plutonium obtained using pyrochemical reprocessing methods. Additional residual power release of fission products and minor actinides as well as neutron emission of minor actinides being a part of vibro-packed MOX fuel composition do not influence significantly the process of reprocessed MOX fuel management. TABLE 3.13. MAIN CHARACTERISTICS OF GRANULATED FUEL PRODUCED USING PYROCHEMICAL METHODS
Characteristic Kind of granulated fuel
UO2 PuO2 (U, Pu)O2 Metal content, % of mass 87.8−88.0 87.7 87.75 Pycno-metric density of granules, g/cm3 10.7−10.8 10.9 10.7 Packed density of poly-disperse granulate, g/cm3 6.0−6.2 4.0 6.0
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Density after vibro-packing, g/cm3 8.8±0.2 - 8.8±0.2 Oxide/metal ratio (oxygen coefficient) 2.00+0.01 - - Technological impurity content: mass share of chlorine–ion, % of mass 0.007 0.005 0.006 mass share of carbon, % of mass 0.003 0.002 0.006
TABLE 3.14. CONTENT OF MA AND FP IN PYRO-CHEMICALLY REPROCESSED FAST REACTOR MOX SF Share of original content in regenerated Chemical element material after reprocessing Uranium (all isotopes) 0.997 Plutonium (all isotopes) 0.993 Neptunium 0.5 Americium 0.2 Curium 0.05 Zr, Nb, Ru, Rh, Pd, Tc 0.05 Cs, Rb 3.33 . 10-4 Noble gases (Kr, Xe) < 1 . 10-6 Mo 2 . 10-3 J, Br < 1 . 10-4 Ag, Sr, Ba 0.01 Ce 0.1 Other rare earth elements: Y, La, Pr, Nd, Pm, 0.025 Sm, Eu, Gd, Tb, Dy, Ho, Er, Tu, Yb, Lu Other fission products: Sb, Te, Ge, Cd, Se, etc. 0.02
TABLE 3.15. NEUTRON-PHYSICS CHARACTERISTICS OF BN-800 TYPE REACTOR DEPENDING ON MOX FUEL PURIFICATION EFFICIENCY Purification 100 % Purification 0 % Pyro Decontamination coefficients Np ∞ 0 2 Am ∞ 0 5 Cm ∞ 0 20 Effective neutron 1.00391 1.00411 1.00318 multiplication factor Sodium void reactivity effect -0.00115 -0.00084 -0.00116 Doppler-effect 0.00677 0.00667 0.00673 Efficiency of control rods 0.04583 0.04555 0.04570
3.6.2. Incentives and boundary conditions for deployment of INS based on CNFC−FR
Currently the Russian program for the development of the energy sector is being revised with the objective to double annual electricity output by 2020. In the Russian Federation, about 50 % of the total electricity is generated by burning natural gas, while coal provides about 18 %, and the share of nuclear electricity is 16 %, equal to that
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generated by hydroelectric power plants [19]. The big share related to gas results in a decreasing reliability and long-term stability of power supply for consumers. Besides, the further large-scale use of gas in the power industry of the Russian Federation is complicated due to the increasing imbalance between gas production, and its internal consumption and export. The Russian Federation may face gas deficiency as early as in the first quarter of this century. As a result, the need arises for new NPPs. Nuclear power in the Russian Federation is based on thermal reactors – VVER and RBMK, which use 235U as main energy source. The system advantages of nuclear power as compared to fossil fuel are the following: • A world level of commercialized technologies, including VVER reactors and related technologies of nuclear fuel cycle; • A weak dependence on the fluctuations of fossil fuel and natural uranium market prices; • A negligible level of greenhouse and other hazardous gas release; • A low demand for the transport infrastructure. On the other hand, there are some system disadvantages of nuclear power based on thermal reactors using an once-through fuel cycle: • Incomplete implementation of the spent nuclear fuel management cycle; • Unavoidable storage of continually increasing quantities of spent nuclear fuel (over 15 thousand tons of spent nuclear fuel have been accumulated to date, with a rate of accumulation of 800 tons a year); • Accumulation and storage of uranium enrichment tails (the accumulation rate being of ~ 4 thousand tons a year); • Need of a large-capacity repository; • Limited resources of 235U (Fig. 3.45). 235U resources in the Russian Federation are sufficient for achieving a 25 % share of energy produced by NPPs by 2030, but this share would inevitably decrease by the end of the century. At the same time the Russian Federation has enormous strategic resources of 238U, an energy potential which cannot be used with current technology. As can be seen in Fig. 3.45, the energy potential of the 238U resource is 10 times higher than that of coal and 25 times higher than that of natural gas. With the objective of ensuring energy security in the Russian Federation, the nuclear power development program envisions switching by the middle of the 21st century to a new technological platform based on the closure of the nuclear fuel cycle with fast reactors and hence the use of 238U.
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U-238 - 86.7%
Coal - 8.7% Gas - 3.4% U-235 - 0.4% Oil - 0.8%
FIG. 3.45. Relative energy potential of natural resources in Russia.
3.6.3. National planning of capacities and structure of energy sector with determination of the share of nuclear power and fast reactors
Principal directions of nuclear power development in the Russian Federation being an integral part of Russian energy system are defined by the energy strategy of Russian Federation for a period up to the year 2020, called the strategy of nuclear power development in the first half of the 21st century, and by the initiative of the president of the Russian Federation establishing the international project “energy supply for sustainable development of mankind, cardinal solution of nuclear weapon non-proliferation problems and ecological enhancement of the earth”. According to the energy strategy of the Russian Federation up to 2020, electricity generation by nuclear power plants should be increased from 140 billion kWh in 2002 to 195 billion kWh in 2010 and up to 300 billion kWh in 2020. The strategy also envisions a nuclear heat production increase up to 30 billion Gcal/a. Under these provisions, the share of electricity generation by nuclear power plants will rise from 16 % in 2000 up to 23–25 % by 2020 (up to 32–35 % in the European part of the Russian Federation). The analysis carried out by ROSATOM organizations demonstrated that in order to reach the aforementioned targets, about 40 GW(e) of new installed nuclear capacity should be introduced by 2030, resulting in a commissioning pace of about 2 GW(e) per year. Taking into account planned lifetime extensions of existing operating nuclear power units by 15 years (average), the total installed capacity of nuclear power plants in the Russian Federation by 2030 should reach 51 GW(e). Given a preserved commissioning pace of 2 GW(e)/year, the integral installed capacity of Russian nuclear power will make up 81 GW(e) by 2050. At the same time, it is recognized that successful implementation of the targeted commissioning pace of nuclear capacities is possible only through development of a multi- component structure of nuclear power. Directions of high priority in this case will be construction of new reference nuclear power units based on improved “third generation” thermal reactor technologies, and creation of a new technological platform of Russian nuclear power based on fast reactors to provide a long-term sustainable development (Fig. 3.46).
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120
100
80
60
40
Installed Capacity, GWe 20
0 1970 1980 1990 2000 2010 2020 2030 2040 2050
VVER-440 RBMK-1000 VVER-1000
NPP-2006 FR (1,2 GWe/year) FR (3,6 GWe/year)
FIG. 3.46. Russian nuclear power capacity (retrospective view and forecast).
3.6.4. The optimistic and moderate scenarios for development of a closed nuclear fuel cycle with fast reactors
The effect of closing the nuclear fuel cycle and introducing fast reactors into the system of nuclear power becomes significant only at long-term intervals when cheap uranium resources get exhausted. Thus, a scenario study has been carried out with a modeling horizon until 2100. Plausible variants of Russian nuclear power development outlined in the strategy of nuclear power development in the first half of the 21st century have been taken as a basis. Several possible options were selected for the Russian nuclear power structure analysis: • Thermal reactors operating in once-through nuclear fuel cycle (OT-NFC); • Multi-component nuclear system involving uranium VVERs, fast reactors, and MOX fuelled VVERs. Calculation results for the developed models are presented on Figs. 3.47 to 3.50. VVER-1000 with a OT-NFC It can be seen from the results presented in Fig. 3.47 that this “business-as-usual” option requires increasing continuously both the uranium production and separation work capacity. By 2100 almost 1300 thousand tons of uranium will be needed in total, with an increase of annual natural uranium production up to 21 000 tons/a and a rising separation capacity up to 21 million SWU/a.
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FIG. 3.47. VVER-1000 option with an OT-NFC (“business-as-usual”): installed/decommissioned capacity of reactors by type, natural uranium consumption, and separation work.
VVER-1000 with an OT-NFC + FR (breeding ratio ~ 1.05) Introduction of fast reactors with a breeding ratio of ~1.05 (break-even core) by 2010 results in an integral natural uranium consumption of approximately 760 thousand tons, an annual uranium production increase of up to 8000 tons, and a need for separation work of 8 million SWU a year (Fig. 3.48). Thus, a need to considerably increase natural uranium production and enrichment capacities is still retained.
FIG. 3.48. VVER-1000 with an OT-NFC + FR (breeding ratio ~ 1.05): Installed capacity, natural uranium consumption, separation work, and reactors by type. VVER-1000 with an OT-NFC + FR (breeding ratio ~ 1.3) Installation of fast breeder reactors with a breeding ratio of ~ 1.3 by 2010 allows to decrease integral uranium consumption over the modeling interval down to approximately 500 thousand tons, with an annual natural uranium production of just 6.500 t/a by 2020 and
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reduced to 2.500 tons by 2100. Separation work has a peak value of 5.5 million SWU/a by 2020 further decreasing down to 2 million SWU/a by 2100 (Fig. 3.49).
FIG. 3.49. VVER-1000 with an OT-NFC + FR (breeding ratio ~ 1.3): Installed capacity, natural uranium consumption, separation work, and reactors by type VVER (UO2) + FR (breeding ratio ~ 1.6) + VVER (MOX) This option also requires increasing the uranium production and enrichment, though annual performance characteristics rise only during a relatively short time period, decreasing almost threefold after 2020 until 2100. VVER-1000 with an OT-NFC + FR (breeding ratio ~ 1.6) Closing of the nuclear fuel cycle for thermal reactors of VVER type which becomes possible due to a high breeding ratio ~1.6 of fast reactors starting from 2010 (Fig. 3.50) results in an integral natural uranium consumption of about 400 thousand tons, and annual uranium production increases up to approximately 5 500 tons by 2020 and then decreases drastically down to 400 tons a year by 2100. Required enrichment capacities are about 4.5 million SWU/a by 2020, with an abrupt decrease down to almost a negligible value by 2100.
FIG. 3.50. VVER (UO2) + FR (breeding ratio ~ 1.6) + VVER (MOX): Installed capacity, natural uranium consumption, separation work, and reactors by type
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The following conclusions could be made based on the scenario modeling results obtained: • A necessary condition of developing a nuclear power as a system meeting the principles of sustainable development is the perfection of its existing elements, filling its structure with new components and technologies enabling nuclear power to solve its own problems and to facilitate supplying energy in a sustainable manner; • Large-scale nuclear power as a basis for sustainable development should incorporate various types of nuclear reactors and a closed nuclear fuel cycle enabling efficient use of actinides with minimal losses and waste generation; • Fast reactors allow adding valuable flexibility into a nuclear power system due to their inherent ability to breed fissile materials that can be used again as a fuel. At the same time it becomes obvious that there is no single type of fast reactor or single package of fast reactor performance characteristics that would satisfy all national specificities once and for all. 3.6.5. National approach to identifying fast reactors designs and relevant fuel cycles In identifying the concept and R&D program for a fast reactor system with a relevant closed fuel cycle, specific attention should be paid to various factors including the relative importance and topicality of resource and waste issues of existing nuclear power, gained knowledge and experience, and expected additional expenditures related to implementation of fast reactor programs. The Russian Federation, like the majority of the countries developing fast reactor technologies (such as France, Japan, India, China, and the Republic of Korea) adheres to the fast breeder reactor concept that allows to fully address problems of contemporary nuclear power regarding resource availability and waste management. Implementation of a fast reactor technology development program is a strategic task of global character. It requires considerable resources and human efforts to develop, demonstrate and commercialize interconnected elements of the system: fast reactor facilities and technologies of a closed nuclear fuel cycle including reprocessing of thermal reactor spent fuel, fuel fabrication for fast reactors, reprocessing of their own spent fuel, and radwaste management. The problem of defining criteria of effectiveness along with system requirements on the fuel balance of fast breeder reactors in the Russian Federation has a semi-centennial history. In the 1960−1970ies it was generally recognized that the only important criterion was a minimal doubling time, in the period from the 1980ies to 1990ies different values of required breeding ratio (~1.0−1.5) along with a complex of system requirements related to spent fuel management and non-proliferation have been elaborated. As it was shown in the previous sections, limitation of natural uranium resources or enrichment capacities can raise the need for an even higher breeding ratio value. Today in the Russian Federation and in the world, like in 1970ies, an interest recommenced to a large-scale nuclear power development. Again, like in the past, a lot of fast reactor concepts and designs as well as fuel cycle technologies are being proposed, possessing various potential benefits and requiring vast expenses for RD&D and commercialization. Importance of defining proper criteria of efficiency and system requirements for fast reactor systems under such conditions cannot be overestimated. Ambiguity of requirements being posed on fast reactors with respect to fuel resources is connected with the fact that those requirements are usually formulated based on the balance of integral consumption and estimated resources of natural uranium for a long-term (50 years and more) period, until fast reactors get a substantial share in the structure of nuclear power.
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Additionally, it is frequently assumed for simplicity that all fast reactors to be commissioned during this period will have the same performance characteristics. It seems more appropriate to consider development of an innovative fast reactor system during a long term interval as a sequence of stages, for example: • R&D on fast reactors (FR) and related closed nuclear fuel cycle (CNFC) technologies; • Demonstration of a first of a kind FR and CNFC technologies; • Commercialization of FRs and CNFC technologies by constructing a small series of commercial FRs and corresponding “first generation” CNFC facilities for fresh fuel fabrication and spent fuel reprocessing; • Large-scale commissioning of “second generation” FRs and relevant CNFC facilities; • Asymptotic development of nuclear power based on FRs and corresponding CNFC facilities using “third and next generation” technologies. Within the framework of such an approach, the process of INS introduction and deployment could be simulated more realistically providing decision making with a more reliable forecast.
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CHAPTER 4 COMMON AND SPECIFIC FEATURES IN NATIONAL APPROACHES
4.1. EVOLUTION OF TARGETS FOR CNFC−FR DEVELOPMENT
An evaluation of national long term (50−100 years) nuclear energy strategies of the countries participating in the Joint Study indicates a lot in common in the vision of the future of an INS based on CNFC with FR in these countries. At the same time, there are specific national conditions for the strategies’ realization that predetermine rather essential differences in the expected characteristics of the INS. When assessing an INS, it is necessary to make a judgement on both the generic features of the system as well as on national peculiarities based on local needs. It is reasonable to begin the analysis of common and specific features in the development of an INS based on CNFC−FR by a consideration of the evolution of views on its destination and main targets that are the driving forces for changes in design and technology. The system based on a closed nuclear fuel cycle with fast reactors (CNFC−FR) has gone a long and complex way of development. Shortly this way was reflected in the previous chapter of the report. The main features of the system have been formed at the end of the 50ies of the last century. A fast reactor cooled by sodium (SFR), a three- circuit heat transport system, a steam turbine for electricity generation were considered as essential components of a power unit of this type. A pellet technology of fuel fabrication as well as aqueous methods of plutonium extraction mastered by the time completed the vision of a system based on CNFC−FR. Accordingly, at the initial phase of fast reactor technology development objectives of national programs had a lot in common. They were mainly focused on demonstration of: • Breeding capability of a fast reactor; • Technical feasibility and operational reliability of the system; • Safety; • Economic competitiveness. The results gained by the middle of 90ies in development and operation of fast reactors and fuel cycle installations allowed the conclusion that objectives posed at the beginning of technology development had been achieved to a great extent. The breeding capability of fast reactors as a fundamental characteristic for efficient utilization of natural uranium had been fully established with PHENIX and BN-350 fast reactors [20]. The total period of SFR operation had exceeded 200 reactor-years. The time of operation without major repair of the main components of experimental and demonstrational SFR had reached 200 000−250 000 hours (30−35 years). Basic engineering and technological approaches in core physics, coolant technology, structure materials, thermal hydraulics, fuel fabrication and reprocessing, etc., had shown their efficiency in providing technical feasibility and operational reliability of electricity production in the system based on SFR. It was demonstrated at operating units with fast reactors that this technology provided acceptable safety standards [21]. SFR had demonstrated minimal impact on the environment and humans during operation. Heat releases from a fast reactor were essentially lower compared to a thermal nuclear reactor. Gas-aerosol releases, iodine- 131, and long-lived and short-lived nuclide releases of SFR were so low that they
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practically were not detectable, and inert radioactive gas releases were a hundredth fraction of admissible levels and were mainly determined by experimental studies in "hot" chambers. The mastering of accidents that occurred in fast reactors with leakages of water into sodium and sodium leakages proved that engineering methods for safety provision of sodium circuits were available. The economic characteristics of fast reactors prototypes gave grounds to go ahead in refinement of the design and technology with the aim to reach economic competitiveness with thermal reactors and coal power plants. Thus, experience in development, construction and operation of sodium cooled fast reactors and installations of their fuel cycle proved the attainability of the objectives posed at the initial stage of development. However, it appeared that reaching the initial objectives did not lead to a significant expansion of SFR systems in current energy markets. This situation was caused by external and internal effects. At the end of the 70ies energy markets had begun to change not in favour of nuclear power, especially in developed countries which at the time were leaders in the technology development of fast reactors: A firm policy for energy saving led to a reduced growth rate of energy demand; globalization of the oil and natural gas markets and subsequent slump of the cost of organic fuels mitigated the problem of assurance in primary energy supply at national level; increased concern for environment preservation accelerated a revision of the approaches regarding nuclear safety and waste management towards more stringent requirements and costly solutions; and deregulation of electricity markets resulted in decreasing (up to a full loss of) the investment attractiveness of nuclear power plants because of the low level of internal rate of capital return. These significant changes happened in the energy and electricity markets of many countries by the end of the century and revealed deep inner problems of nuclear power and forced them to adjust the main objectives of its development to these new realities. For many reasons the INS based on CNFC−FR appeared at the front end of critical analyses. Among the new requirements there were several ones that implied radical updating of characteristics of fast reactor technologies. A doubt was cast even on the initial destination of the technology – nuclear fuel reproduction. Countries with a mature nuclear power sector based on thermal reactors did not get into trouble for a lack of plutonium. On the contrary, they had to solve ecological problems connected with accumulation of plutonium and minor actinides in spent fuel. A few countries had installed costly storages of separated plutonium which was accumulating without being used by nuclear industry. These realities initiated a concept of a fast reactor ‘burner’ of plutonium and minor actinides – a concept quite opposite to the original concept of a fast reactor ‘breeder’. Further, severe accidents at Three Miles Island and Chernobyl clearly demonstrated the global character of radioactive impact of such accidents as well as the scale of their consequences. New ideas and approaches to assure the safety of an INS based on CNFC−FR were proposed which affected the essence of the established vision of this system. These proposals stirred up a discussion of specialists and stimulated national regulatory bodies and the IAEA in development of renewed safety rules and standards. Economic realities of the 80ies−90ies made it necessary to update also the objectives regarding competitiveness of nuclear power plants. Decreasing governmental support of nuclear power in many countries, significant technical progress in energy technologies
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based on the fossil fuels resulting in improvement of economic indicators, an urge of investors to cut investment risks, – all these factors imposed new stringent economic requirements on innovative nuclear systems, especially on their investment attractiveness. At last, rising concerns regarding nuclear proliferation and threat of terrorism made it necessary to include into the list of main requirements on technologies and organisation of a fast reactor fuel cycle new important issues of proliferation resistance and physical protection which were not included into initial objectives of technology development. A large number and radical character of challenges arose for fast reactor technology development in the 90ies; this indicated that the important phase based on the initial set of objectives had ended. It became obvious that new prerequisites for a new phase of technology development had been created and that this new phase should be based on a philosophy tuned to contemporary realities and predictable requirements of the future. The need in creation of a methodological platform which would imbibe and harmonise new ideas and requirements for a long-term nuclear power development was clearly realized by the nuclear community. The task has become an integral part of two international projects initiated in the year 2000: Generation-IV International Forum (GIF) launched up by the US Department of Energy [22] and the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiated under the auspices of the IAEA [1]. The concept of sustainability and sustainable energy supply was referred to as an overall guide in both projects. For the time being the Generation-IV and INPRO projects are being coordinated with the aim to achieve identity in the main objectives of INS development. A set of basic principles, user requirements, and criteria following the concept of sustainable energy supply was defined in the INPRO methodology [2], [5] and agreed at an international level for each area of concern (economics, safety, environment, waste management, infrastructure, proliferation resistance and physical protection). Participants of the study considered the principles and requirements formulated in INPRO documents as a foundation for their work and do believe that they constitute a firm, albeit flexible, basis for a new phase of development of an INS based on CNFC−FR.
4.2. OBJECTIVES, MILESTONES AND TIME FRAMES OF NATIONAL PROGRAMMES
One of the key challenges to sustainability of the world development is to find a way how to satisfy increasing energy needs, including such needs in the most populated and dynamically developing Asian region, without having catastrophic environmental and health impacts. Many studies indicate that global sustainability goals will not be achieved with present patterns of energy consumption and productions. Innovations of the technological basis in different parts of the energy sector would be essential for the success. The countries participating in the Joint Study are among the biggest consumers of energy resources in the world. The regional energy sector under consideration in the Joint Study is enormous, with a vast total amount of power generated by all kind of industrial energy technologies. As stated before, the potential of nuclear power deployment in the region of the Joint Study amounts to 400−600 GW(e) by 2050.
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Nuclear power evokes rising expectations as a prospective, though disputable, option for such innovation. A weak dependence on fluctuations of market prices of natural uranium, a negligible level of greenhouse and other harmful gas releases, a small environmental and health impact during normal operation, and a low demand for transport infrastructure – all these and some other fundamental properties of nuclear energy provide a potential for making existing energy supply systems more sustainable. In the countries participating in the Joint Study it is anticipated that in a future energy mix nuclear energy would mitigate some problems connected with energy sources based on fossil fuels (resource depletion, instability of oil and gas prices, significant greenhouse emissions, considerable health impact, etc.). In turn, a two component system of thermal and fast reactor is being considered as an option for providing stable performance of nuclear power: thermal reactors would feed fast reactors with plutonium reprocessed from their spent fuel, while fast reactors would contribute to improving the characteristics related to sustainability of nuclear fuel cycle by means of more efficient use of natural uranium, reduction of the amount of uranium enrichment tails, reduced size of storages for spent fuel and large-capacity repositories, enhancement of proliferation resistance, etc. Thus, the intention to complete development of an INS based on CNFC−FR as an integral part of a future sustainable energy mix is an important common feature in the strategies of nuclear power development discussed in Chapter 3. There is also a lot in common in understanding the milestones of CNFC−FR system development: All participants of the Joint Study are going to continue development and demonstration of INS based on CNFC−FR in a transition phase to significant deployment of CNFC−FR. The first step of the transition phase will be finished by installation of a pilot commercial fast reactor with assured characteristics of economic competitiveness coupled with fulfillment of the acceptance limits of sustainability in the area of safety, environment, waste management, proliferation resistance, physical protection and infrastructure. In the second step of the transition phase after a successful operation of a pilot commercial fast reactor is demonstrated, a series of these reactors are supposed to be deployed. A steady growth of installed nuclear capacity based on improved thermal reactors would continue during the transition period. Utilization of spent fuel of thermal reactors in fast reactors (closure of the fuel cycle of thermal reactors) will become a significant task. The length of the transition period to a large-scale deployment of CNFC−FR differs from country to country (participating in the Joint Study) but remains in a timeframe of 15−30 years. Specific objectives to be achieved during this period are: • Commercialization of the components of CNFC−FR based on a sodium fast reactor and demonstrated fuel cycle technologies; • Development of an optimal strategy for transition to an INS based on serial thermal and fast commercial reactors with elements of industrial closure of the nuclear fuel cycle; • Elaboration of a vision of a mature INS based on CNFC−FR as a global nuclear energy system of 21st century, determination of requirements on its components in the area of safety, economics, environment, proliferation resistance, and infrastructure; • Identification of directions of R&DD for development of INS based on CNFC−FR both for the sodium fast reactor system and for a systems based on fast reactors with other coolants.
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The following basic tasks should be solved during the transition period: • Experimental studies and tests for mastering fast reactor fuels, refinement of reprocessing technologies and technologies of waste disposal using existing or newly commissioned experimental and prototype reactors and installations of the nuclear fuel cycle; • Completion of design and commission of a pilot commercial fast reactor; • Creation of the infrastructure for deployment of the INS based on CNFC−FR including a system of training specialists and skilled workers, construction of pilot commercial plants for reprocessing fuel of thermal and fast reactors and for fabrication fuel for fast reactors, extension of the industrial sector for fabrication of instruments and equipment; • Commission of a small series (3‒5 units) of commercial fast reactors; • Carrying out R&DD on advanced designs and technologies of CNFC−FR providing break-through characteristics of sustainable energy supply: o Innovative concepts of fast reactors; o Prospective fuels for fast reactors and advanced technologies for fuel reprocessing; o Geological facilities for the final isolation of radioactive waste.
4.3. SCOPE OF DESIGNS AND TECHNOLOGIES UNDER CONSIDERATION
A brief summary of the current national approaches to select the designs and technologies for development of fast reactor technology and associated closed fuel cycle is presented in Table 4.1. It follows from Table 4.1 that the sodium cooled fast reactor is being considered by a majority of the Joint Study participants as the most mature fast reactor option that has a significant possibility in the countries to meet INPRO requirements of sustainability, including economic ones, within the next 15−25 years. Reactor units of installed capacity above 500 MW(e) based on an evolutionary improved three-circuit scheme are expected to be dominating during that period. Pellet technologies for MOX fuel fabrication as well as advanced aqueous and pyroelectrochemical technologies for reprocessing are the most probable candidates to be used in this period. In the Russian Federation the vibro-packing technology of fuel fabrication is also mastered and demonstrated at BOR-60 and BN-600 fast reactors. This technology has real prospects for commercial introduction in the near future.
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TABLE 4.1. NATIONAL APPROACHES TO THE DEVELOPMENT OF AN INS BASED ON CNFC−FR
Fabrication/ Stages of FR development and reactor configuration Coolant Fuel reprocessing Some specific features of Countries Demo Commercial Ref. Var. Ref. Var. Ref. Var. a long term approach Experimental prototype size CEFR, Pellet/ UOX Injection High breeding, in-site fuel China construction, 2020-2025 2030-2035 Na MOX advanced Metal casting/ cycle, FR-burners &ADS pool type aqueous Phenix 1974, Super-Phenix Rapsodie Pellet/ Break even cores, pool type 1986-1998, Car- Plates (for France 1967-1983, Na Gas MOX advanced MA incineration, new FR pool type bide Gas FR) loop type aqueous hydrogen production 2020+ serial 2030+ First serial FBTR 1985, PFBR Pellet/ Sol-gel High breeding, co- 2020, India operation, Construction Na MOX Metal advanced microsphere location of FC facilities, Commercial loop type pool type aqueous pellets/pyro thorium fuel before 2020 Monju 1994, loop type First serial Pellet/ Joyo1977 Loop type reactor Japan JSFR demo 2040-2050 loop Na MOX advanced operation, loop type configuration 2025 type aqueous loop type 2030 assumed Pool type Break even cores, Rep. of Korea Na Metal /Pyro for Joint reduction of SNF Study BR5/10 BN600, 1980 USSR/ 1959-2003, BN350* UOX; BN800 Pellet/ Vibro- Heavy metal coolants, Russian loop type 1973-1999 construction, Pb-Bi, Na MOX Nitride advanced packing/pyro breeding and break even Federation BOR60, 1969, UOX, pool type Pb aqueous cores, modular designs loop type loop type serial 2020- 2025 * Sited at territory of Kazakhstan
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Improved economics, enhanced safety and proliferation resistance, efficient use of U, spent fuel and minor actinides reduction are the common fundamental objectives of the Joint Study countries developing CNFC−FR. The approaches of the countries participating in the Joint Study to select CNFC−FR concepts and technologies become more diverse during the first stage of the transition period. Country-specific novel designs of fast reactors based on new concepts directed to enhance inherent safety, economic, waste management, and proliferation resistance characteristics are being developed (Table 4.1). As a possible alternative to sodium coolant for fast reactors, heavy metal coolants (lead-bismuth, lead) are being examined in the Russian Federation, and gas coolant in France. For developing fast reactors of a modular type national strategies follow different approaches regarding expediency, timeframe and selection of coolant and design features. The USA and the Russian Federation were pioneers in designing such reactors which became a promising direction of nuclear power innovation [23]. The modular systems have potential advantages regarding safety, construction time, proliferation resistance, etc. but they have to be demonstrated as competitive systems to achieve consumer acceptance. As it was shown in the previous chapter, in the Russian Federation a two-circuit concept based on lead-bismuth coolant and conventional steam turbine has been developed to a stage of demonstration. Another innovative two-circuit concept based on sodium as a primary coolant and gas turbine is at an earlier stage of development. Also, modular fast reactor systems are at a conceptual stage of design in some other countries. With a consensus on the necessity to meet requirements of sustainability as defined in the INPRO methodology, the countries participating in the Joint Study have diverse approaches in tuning the CNFC−FR to national conditions. Remarkable physics of fast reactors and a variety of options for closure of the nuclear fuel cycle make it possible to adapt the INS CNFC−FR to specific national conditions and realize diverse aspiration of member states. A drastic reduction of reactor waste requiring repository disposal by reducing the amount of uranium, plutonium, and minor actinides in the waste to be disposed as well as more efficient utilization of uranium and plutonium are the main incentives for developing an INS in the countries with a considerable nuclear share in their electricity sector and a low or moderate expected growth of nuclear capacity (France, Japan, and Republic of Korea). In contrast, fuel assurance through achieving high breeding in fast reactors is a driving force for development of an INS in countries projecting fast growth of nuclear power from a low level of installed nuclear capacity (China, and India). For the Russian Federation where a high growth of nuclear power is being planned to be continued from a rather significant level, both waste reduction and fuel assurance are important reasons for deployment of CNFC−FR. The national preferences for the destiny of CNFC−FR system determine the selection of innovative fuels. Nitride fuel is being considered in the Russian Federation as an appropriate choice for sodium and lead coolant fast reactors with enhanced inherent safety characteristics [24]. High-dense metallic fuel as an option for providing high breeding ratio in fast reactors is selected by China, India and Republic of Korea to meet national needs of nuclear fuel assurance in programs of a rapid growth of nuclear capacities. France is examining carbide fuel. In fact, it is very difficult to predict to what extent national priorities will diversify the characteristics of future fast reactor systems. No doubt, innovative systems will be
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tailored to local needs. At the same time, there are important factors which are working in favour of INS unification. Trends in world globalization compel to find technical solutions to be the most efficient at a global level; international collaboration and trade to an increasing extent compensate the lack of resources or components of infrastructure at the national level. It is supposed that some opportunities determined by these trends will be considered as a part of the Joint Study implementation and realized in the future.
4.4. PRECONDITIONS FAVOURING COLLABORATIVE EFFORTS
The countries of the Joint Study represent a wide spectrum of political, socio-economic, and cultural systems. To illustrate the diversity, a dispersion of quality of human life in the countries is shown in Fig. 4.1 as a function of electricity consumption per capita.
1.0 U.N. Human Japan US Development Index France Republic of Korea Canada
Russian Federation China
India
0.3 4000 8000 12 000 Annual electricity use kWh/capita FIG. 4.1. UN’s human development index as a function of electricity consumption. The history and status of nuclear power in the countries participating in the Joint Study are also rather different. A multi-regional analysis [25] indicated a potential facilitating growing nuclear deployments due to synergies between regions, reactors and fuel cycles. The Joint Study gives an opportunity to specify major benefits of international collaboration in the area of CNFC−FR technology development and to determine the feasibility of harmonization of national programs with regional and world expectations. The expected growth of total and nuclear energy generation in the countries participating in the Joint Study is not distributed evenly. Specific national macroeconomic and social conditions determine different expectations for the growth of energy demand: from about ‘zero’ up to ‘as much as possible’. Level of quality of life is one background that determines these significant differences in energy strategies. Dependence between the national level of electricity consumption per capita and the United Nation’s Human Development Index [10], [26] presented in the Fig. 4.1
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demonstrates that in spite of the fact that China and India are among the leaders in total electricity production in the world, the electricity consumption per capita in these counties is much lower than in Canada, USA, Japan, France, and the Russian Federation. China and India need a capacity addition of about 10 GW/year to support GDP growth in quest for a better quality of life. On the contrary, Canada, Japan, and France are very close to reaching the level of saturation in energy consumption. Their expectations for further increasing the quality of life are connected with safe, secure and ecologically clean energy supply rather than with its growth. Declared nuclear deployment plans for 15−25 years indicate significant unevenness in the expected growth of nuclear capacities. Table 4.2 illustrates near- and medium-term nuclear plans of some countries participating in the Joint Study (with reference to data presented in Ref. [27]). TABLE 4.2. NEAR-TERM NUCLEAR PLANS IN SOME COUNTRIES OF THE JOINT STUDY
Country of Current capacity in 2005 Expected capacity Expansion the Joint Study GW(e) GW(e) (electricity share, %) (electricity share, %)
(Asia) China 6.6 (2.3) 40 (4) by 2020 x 6 India 3.0 (2.8) 29.5 (10) by 2022 x 9 Japan 46.5 (29) 58 (30-40) by 2030 x 1.2 Rep. of Korea 16.8 (39) 26.6 (34) by 2015 x 1.6
(Europe) France 63.1 (78) 63 by 2025 x 1 Russian Federation 21.7 (16) 35 (25) by 2025 x 1.5 Ukraine 13.1 (48.5) 20-22 by 2030 x 1.6 National forecasts presented in Chapter 3 show that trends reflected in Table 4.2 are expected to be preserved for a longer time horizon as well. These forecasts also envisage growth of fast reactors in national nuclear fleets. To realize these expectations, a mature infrastructure of a closed fuel cycle has to be created; however, the countries participating in the Joint Study are in diverse initial positions to do this because of their different nuclear history. In the 50ies till the 80ies of the last century several countries including the USA, the USSR, UK, France, Japan, and Germany had vigorous fast reactor development programs. A wide spectrum of experimental installations for mastering the main stages of technology for closure of the nuclear fuel cycle had been created in all these countries. In the 80ies, at the peak of the nuclear activity, the industrial infrastructure for the three main stages of the fast reactor technology, i.e. uranium-plutonium fuel fabrication, fast reactor operation, and spent fuel reprocessing had been established. Two industrial size fast reactors (BN-350, BN-600) were operating by the time in the USSR. They demonstrated technical feasibility and operational reliability of reactors of the type as a precondition for creation a complete INS based on CNFC−FR. In 1980 at the productive unit “Mayak”, Chelyabinsk a MOX fuel pellet fabrication facility
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“Paket” was constructed as a pilot industrial facility for the future large scale MOX fuel production [28]. In the research institute of atomic reactors (RIAR), Dimitrovgrad, pilot MOX facilities based on a vibro-compact technology of MOX fast reactor fuel fabrication combined with a pyro-electro-chemical method of fuel reprocessing were constructed and used for fabrication of experimental fuel rods for the BOR-60, BN-350 and BN-600 reactor. At the radiochemical plant RT-1 plutonium was being recovered from nuclear spent fuel, albeit its extensive use in fast reactors was postponed until commissioning of a large scale industrial plant for fast reactor MOX fuel fabrication. A coherent development of reactors, fuel fabrication and reprocessing plants was performed in France. All standard and experimental fuel for fast reactors was fabricated in the Cadarache plutonium workshop. Its capacity was extended from 0.5 ton of MOX/a for a RAPSODIE line to 4 t/a for a PHENIX line, and to 20 t/a for the SUPERPHENIX [28]. Fuel from French fast reactors was reprocessed in plants at Marcoule and La Hague. In Japan all standard and experimental fuel for fast reactors was fabricated at the three fuel fabrication facilities of Tokai [28]. Basic research for reprocessing of fast reactor spent fuel has been carried out since 1982 at Tokai chemical processing facility. In the 80ies deployment of nuclear fuel cycle infrastructure for an INS based on CNFC−FR in countries which originally set up the ‘fast reactor club’ began to slow down, up to a full stop in the UK and Germany. At the same time, some countries, France in Western Europe and Japan in Asia, continued to develop the infrastructure for the use of MOX fuel in light water reactors. Currently, this option has reached a high level of maturity. World-wide MOX fuel fabrication capacities exceed 500 t(HM) per year [29]. Reprocessing capacity amounts to 5000 t(HM) per year and will increase with the start-up of the new plant Rokkasho-mura with a capacity of 800 t(HM)/a [30]. Accumulated knowledge and expertise as well as some components of the infrastructure created for the management of MOX fuel in light water reactors will be helpful in case of expanding the use of MOX fuel in fast reactors. While the activity of some old members of the ‘fast reactor club’ regarding development of INS based on CNFC−FR declined, a real need for growth of energy supply made China, India and Republic of Korea active proponents and developers of fast reactor technology. By now important segments of infrastructure for the CNFC−FR are created in these countries including an education and training system, research centers, and an experimental, design and construction base. India is one of the countries to have developed expertise in the nuclear fuel cycle. The Bhabha atomic research centre (BARC) and the AFFF plant in Tarapur manufactured MOX fuel assemblies for irradiation in thermal reactors. BARC also fabricates mixed carbide sub-assemblies for the Indian fast breeder test reactor (FBTR) and MOX test fuel assemblies for PHWR and the Indian prototype fast reactor (PFBR) [31]. Prehistory of nuclear power development in a country is, probably, especially important for fuel assurance during fast reactor deployment. Plutonium as the main fissile material to be used in the fast reactor fuels practically does not exist in nature and is being produced in sizeable quantities only in nuclear reactors of different types. The plutonium content of spent fuel, which depends primarily on burnup of the fuel, is typically 6–12 g/kg of uranium for light water reactor fuel (~1 %) [29]. Existing reprocessing technology is efficient, with greater than 99.5 % of plutonium being recovered [32].
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The total amount of spent fuel generated by the end of 2000 is expected to be close to 230 000 t(HM) (worldwide) [28] along with about 160 t of separated plutonium [29]. Projections indicate, that the amount of spent fuel generated by the year 2020 is expected to be in the order of 445 000 t(HM) and the amount of separated plutonium to be slightly decreased to approximately 120 t. Generally speaking, this separated plutonium and plutonium accumulated worldwide in spent fuel make a sufficient reserve for deploying a rather large fleet of fast reactors. But the main amounts of spent fuel as well as of separated plutonium are accumulated in countries which had an early start of their nuclear programmes based on wide deployment of thermal reactors. More spent fuel and plutonium is being accumulated than is being used in these countries. As a result, they are facing a problem of storage, utilization or disposal of increasing amounts of spent fuel. Some participants of the Joint Study: France, Japan, Russian Federation, and Republic of Korea belong to this group of countries. Strategies which are being realized by France and Japan in respect to spent fuel inventory have much in common. These countries do not need a fast growth of nuclear energy in the future whereas the rate of spent fuel generation in these countries is the highest among participants of the Joint Study. To use the energy potential of Pu accumulated in spent fuel and reduce spent fuel inventory they have resorted to the option of utilization of reprocessed Pu in MOX fuel for existing light water reactors. This stage is being considered as a transitional option to the more remote stage when reprocessed Pu will be used in fast reactors. The Russian Federation and Republic of Korea which foresee higher growth of nuclear energy capacities than France and Japan adhere to another approach: It intends to escape the stage of MOX fuel utilization in light water reactors by moving directly to a stage of reprocessed plutonium utilization in fast reactors. As an exception, the nuclear strategy of the Russian Federation admits utilization of weapons plutonium, in the amounts determined by political agreements, in the form of MOX fuel for light water reactors (VVERs) reactors. An important common feature of these four countries (France, Japan, Republic of Korea and the Russian Federation) is the fact that national programmes of fast reactor development are assured by necessary reserves of the fuel material. Owing to this, there is no need for them in a short- and even in medium-term prospect to speed up the breeding capability and doubling time of fast reactors. For China and India which have not accumulated at the moment large amounts of Pu in spent fuel but have a very high need of growth of nuclear energy, the issue of thrifty use of Pu is a key issue of the national nuclear strategies. These two countries of the Joint Study have oriented their strategies to the use of reprocessed Pu only in fast reactors with an intention to reach the highest possible characteristics of fuel breeding as soon as possible. Results of the consideration of some factors effecting the development of CNFC−FR in the countries of the Joint Study are compiled in Table 4.3.
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TABLE 4.3. COMPLEMENTARITY AS A PRECONDITION FOR COLLABORATION
China France India Japan Republic Russian Optimal of Korea Federation conditions for FR development Energy Very Low Very high Low High High Very high demand high Small, High, Very high, Nuclear Very high, Small, fast High, Very high, fast stabili- fast share stabilizing growing growing growing growing zing growing FR Experi- Commer- Design Commer- Commer- technology mental Demo Demo cialization works cialization cialization maturity program Highly Developed, Highly developed, Highly Highly CNFC FR carbide deve- FR UOX developed, developed, Techno- Up- and MOX loped, Up- fuel FR MOX UOX and logy coming fuel FRMOX coming mastered, fuel MOX fuel maturity demon- fuel FR MOX mastered mastered strated mastered demon- strated Pu in SNF Small High Small High Moderate High High The data compiled in the Table 4.3 reflect a remarkable situation: no country of the Joint Study, taken separately, shows all factors favouring development of the CNFC−FR to a maximal degree but the system in the whole does. Low increment in energy demand in some countries under consideration would hamper utilizing to a full extent the potential of their developed CNFC−FR infrastructure while lack of CNFC−FR infrastructure in the countries with a vast increment of energy demand would constrain their growth of nuclear power deployment. All states participating in the Joint Study envisage further development of their national nuclear infrastructure for CNFC−FR. However, in the countries with high maturity of their nuclear infrastructure − constrained by a stabilized capacity of national energy markets − an incentive for further development of the CNFC is expected to be very limited, whereas a strong orientation towards domestic resources and infrastructure in countries with a high demand for nuclear growth might lead to a rather strained situation regarding nuclear fuel supply. The above overview indicates generic preconditions for reaching mutual benefits thanks to multilateral approaches to the nuclear fuel cycle [33]. With time, specific situations fixed in Table 4.3 may change but the expediency in sharing infrastructure and getting services of industrial and finance entities at a regional and global level will undoubtedly become more urgent owing to the inevitable growth of the number of nuclear energy users, operations in the fuel market and intensification of fuel flows. For the time being, a healthy world market exists only at the front end of the nuclear fuel cycle of thermal reactors. Extension of regional/global nuclear services to the back end of the nuclear fuel cycle could be put into practice under the indispensable condition that fissile materials of spent fuel would have had an opportunity to be reused as a valuable and economically attractive energy resource. Being one of the most promising options for efficient utilization of fissile materials from spent nuclear fuel, the INS CNFC−FR should be considered as a key element of a
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future multilateral nuclear fuel cycle which could provide a full spectrum of nuclear fuel services and become an essential supplement to a national nuclear infrastructure in case when its capacities were not sufficient for satisfying the needs of desirable nuclear deployment.
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CHAPTER 5 GLOBAL CONTEXT OF THE ASSESSMENT
5.1. COMMON VISION AND PARTNERSHIP
At present there is a worldwide consensus that a long-term assurance of energy production, distribution, and use are crucial for achieving the overall goals of social, environmental, and economic sustainability established by the "Earth Summit" in Rio de Janeiro [34]. Several assessment studies of the INPRO project were launched to demonstrate on the basis of INPRO methodology expediency to enhance deployment of nuclear power for fulfilling energy needs in a sustainable manner [1]. Most of them were confined to an evaluation of national nuclear systems. The Joint Study is one of the assessment studies guided by principles and requirements developed in the INPRO methodology. But the scale of the problem formulated in it exceeds national frontiers. Transition to sustainable patterns of energy supply in the countries participating in the Joint Study would make a great contribution to a more equitable, environmentally sound, and economically viable life in a significant part of the world. As it follows from the concept of the study, an INS based on CNFC−FR is considered as a system with the potential to become an integral part of a global solution to a global energy challenge. The considerations made in the previous chapter of the report show that joint collaborative efforts by countries with different social, economic and technological conditions could provide a synergy effect resulting in the acceleration of nuclear power deployment. Eventually, response to the global energy challenge will require a global vision of opportunities and a global partnership. The goal of this chapter is to highlight some issues related to the global context of the assessment. The common standpoint of the Joint Study’s participants is to make a contribution developing a global vision that should reflect a broad and dynamic picture of INS deployment. External conditions in which the INS exists and an internal reserve of the INS’ competitiveness are different from country to country during a given period of time as well as they change in time. Thus, the potential of INS sustainability cannot be described by a constant value: it strongly depends on time, local and global resources and the availability of alternative innovative energy technologies. Simulation of an energy mix taking into account specific local and global conditions gives an opportunity to examine the role of the INS in providing sustainable energy supply in a short-, medium-, and long-term prospect. A top-down approach to INS assessment based on modeling of energy mix and cooperation of the INS’ assessors with experts from energy planning organizations was suggested in Section 5.1 of the Volume 1 of INPRO Manual [35]: • Step1, macro level − determination of the expected growth of population, GDP, energy demand, resource and environmental limitations at a global/regional/national level; • Step 2, energy sector − simulation of an energy mix (fossil, renewables, nuclear) with the aim of identifying its optimal structure (resource and financial balances, requirements on infrastructure), development of a sustainable energy scenario; • Step 3, nuclear power − screening the capability of different nuclear systems to meet requirements of sustainable energy development and determination of a sustainable nuclear energy system within given energy mix (modelling of a nuclear fuel cycle using a holistic approach and making a judgement based on the INPRO methodology).
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After carrying out of these three steps it seems expedient to make a second iteration taking into account results of the implementation of step 3. This step has to add technological details and limitations specific for a nuclear option which might correct the plan of nuclear deployment received at step 2 of the first iteration.
For the time being, comparison of economic indices of nuclear and non-nuclear energy options has become an integral part of the INS assessment procedure using the INPRO methodology [2], [35]. The approach based on a comparative assessment of nuclear and non- nuclear energy options for all dimensions of sustainability was also used in a Russian case study at the stage of INPRO methodology validation. From our point of view, it is reasonable to extend the application of this approach to all topical areas of the INPRO methodology to reach an unbiased evaluation of the nuclear option in a future energy mix. The participants of the Joint Study believe that in the future the INPRO methodology (and its manual) will provide an assessor with necessary details regarding full realization of the approach. Meanwhile, some elements of the above scheme will be used below to show the relevance of a global analysis making a judgment on the potential of an INS to meet the requirements of sustainability.
5.2. DISCUSSION OF SCENARIOS
The national scenarios presented in Chapter 3 have demonstrated some realities that are very important for the evaluation of an INS CNFC−FR at the national level. But having been constrained by the national limitations, every national strategy taken separately gives no opportunity to see the overall picture and to understand global trends. For a better understanding of common problems of nuclear power development and the possible role of fast reactor technology to solve these problems, participants of the Joint Study compared the results of national scenario studies with a more generic global picture. In accordance with a top-down approach, discussed in the previous paragraphs, this consideration should be started from a macro level analysis and provide an assessor of an INS global scenarios of nuclear power deployment. It was decided by the participants of the first consultancy meeting of the Joint Study to address a set of scenarios developed by the inter-governmental panel on climate change (IPCC) and documented in a special report on emission scenarios (SRES) [8]. As known, the SRES scenarios extended to 2100 were developed to provide a common international basis for researchers and policy makers analyzing potential impacts by a climate change. A fundamental approach to macro level global analysis, wide time horizon, and broad scope of energy options under consideration are characteristic features of the IPCC research. These features are very helpful in solving the tasks defined in the INPRO project, therefore some results from the IPCC report have already been used in INPRO technical documents Refs [1], [2], and [35]. The arrangement of geographical zoning in the SRES output data spreadsheets does not envisage the specific region covered by countries of the Joint Study. It makes it impossible to receive a SRES forecast on the nuclear power development in this region. Nevertheless, an expert analysis made in the Joint Study with the aim to focus the forecast − instead on the whole world − on the countries under consideration provided the opportunity to make a rough comparison of SRES global projections with the expectations of the countries participating in the Joint Study. The results of this comparison are shown in Fig. 5.1.
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Range of nuclear power in the countries of JS
3000
2500
2000 JS from A1 JS from B1 1500
GWe JS from B2 1000 JS 500
0 1990 2000 2010 2020 2030 2040 2050 2060 2070 yr
FIG. 5.1. Projections of nuclear capacities for the countries of the Joint Study. Fig. 5.1 illustrates that SRES scenarios built for specific evaluation of the problems related to emissions of greenhouse gases show a too low accuracy for providing a valuable guide for nuclear decision making. Indeed, the high uncertainty of nuclear energy growth in the forecasts based on SRES storylines projected from the whole world to the countries under consideration, would admit realization of any conceivable possibility of nuclear power deployment in the countries of the Joint Study. Under these circumstances, it is impossible to make any reasonable conclusion on the expediency of fast reactor introduction, desirable breeding characteristics of fast reactors and on their optimal share in nuclear power. As shown in Fig. 5.1, a consolidated rate of nuclear power growth after 2040 based on predictions of national nuclear strategies is rather consistent with the development forecasted by the SRES B2 scenario. Similar predictions are also made by the SRES A2 scenario (not shown in Fig. 5.1). Taking into account the important role of the countries in the future development of nuclear energy worldwide, the A2 and B2 global scenarios should be considered as the ones that are to some extent confirmed by national expectations. The A2 and B2 SRES scenarios predict a growth of nuclear power up to 2 000 GW(e) by 2050, and 6 000 GW(e) by 2100. The target values in these scenarios are rather modest compared to maximal nuclear capacity levels predicted by the SRES scenarios of rapid economic growth (e.g., A1 in Fig. 5.1) with an intensive penetration of the world energy market by nuclear power , i.e. 4 000 GW(e) by 2050 and 16 000 GW(e) by 2100. At the same time, the forecast of nuclear power deployment in the A2 and B2 scenarios avoids the nuclear power stagnation projected by some scenarios from the SRES B1 scenario family (see Fig. 5.1). It is not unlikely that national projections can underestimate the long-term prospects of nuclear power deployment, especially the ones related to non-electrical nuclear applications (heat production, hydrogen production, etc.). These non-electrical applications play a significant role in assumptions used in the SRES scenarios with high growth rates of nuclear power. From this point of view, more aggressive global scenarios of nuclear power growth than those suggested by the A2 and B2 SRES scenarios could be considered as a demonstration of the potential of nuclear technology to contribute fulfilling high energy needs in the 21st century in a sustainable manner, while the global A2 and B2 scenarios should most likely be considered as the ones describing an evolutionary mode of nuclear power development.
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5.3. HARMONIZATION OF NATIONAL AND GLOBAL SCENARIOS
5.3.1. Assurance of uranium fuel supply
Assurance of nuclear fuel supply is one of the key requirements of sustainability for global nuclear energy deployment in the 21st century. Fulfillment of this requirement depends on availability of natural fissile material resources and strategies of nuclear power development. In the INPRO methodology a provision related to assurance of nuclear fuel supply is reflected by an environmental user requirement: “The INS should be able to contribute to the world’s energy needs during the 21st century without running out of fissile/fertile material and other non-renewable materials, with account taken of reasonably expected uses of these materials external to the INS.” This requirement is detailed in the INPRO methodology via specific criteria of availability of fissile/fertile materials to be fulfilled by satisfying the corresponding acceptance limits. An example of performing a judgment on the potential, i.e. capacity, of an INS to fulfill the criterion related to availability of natural uranium for a century in the INS is discussed below with reference to a material balance analysis documented in Ref. [36]. Estimations of uranium resource are documented in the “Red Book” published by NEA/OECD as a summary of the world data on reported uranium resources, production and demand [10]. According to the Red Book, uranium resources consist of different categories: well-known uranium resources of the world (called reasonable assured resources (RAR) and estimated assured resources (EAR)) with a price below 130 $/kg of uranium; additional resources’ estimations – based rather on trends and similarity of characteristics; theoretical resources estimated on geological extrapolations; and, finally, resources included by expert considerations without any cost estimations. These categories result in a total of 15.3 million of metric tons of uranium. Inclusion of already extracted resources: commercial reserves, military material stocks, and uranium tails processed with modern technologies of enrichment could add to the estimation of world’s uranium resources about a million of metric tons of material equivalent to natural uranium. This consideration does not exclude a possible increase in the estimation of uranium resources with more active uranium deposits’ exploration. However, a higher estimation of the amount of the uranium available for extraction is rather speculative since it is based either on qualitative considerations, or on extrapolations. Besides, preliminary assessments made by some participants of the Joint Study have shown that the price of uranium fuel fabricated from uranium deposits beyond a value of 16 million tons at a cost above 130 $/kgU could reach a limit beyond which economic incentives of its utilization will not be evident compared to the use of plutonium in a closed fuel cycle. Thus, sixteen millions tons of natural uranium is used below as a reference value of the economically attractive world uranium resource. Thorium can also expand the fuel base of nuclear power several times, but the involvement of fuel based on a thorium fuel cycle requires creating of an industry for extracting and processing thorium, as well as development of technologies for converting thorium into 233U. However, in case high rates of deployment of nuclear power would be limited by availability of cheap uranium, large-scale adoption of thorium could not solve this problem because the doubling time of nuclear fuel in thorium systems is much longer than in systems with fast reactors operating on uranium–plutonium fuel. As it is known, the efficiency of natural uranium utilization in a nuclear system based on thermal reactors is very low. Less than 0.05 % of natural uranium burns up. Approximately 90 % of the mass goes into waste in the form of depleted uranium, about 10 % of it is
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converted into spent fuel. Fig. 5.2 demonstrates a possibility to realize a strategy based on light water reactors and a once-through nuclear fuel cycle with the aim to reach the rates of nuclear power growth in accordance with scenario SRES B2 and at the same time to limit consumption of natural uranium at the level of 16 million tons. The calculations are performed with use of the DESAE computer code [9].
FIG. 5.2. Installed capacity of thermal (blue) and fast (brown) reactors with a maximum consumption of natural uranium of 16 million tons (breading ratio of fast reactors is 1.05). Fig. 5.2 demonstrates that the use of thermal reactors with a once-through fuel cycle (business-as-usual ‘BAU’ strategy) would lead to a very high natural uranium consumption rate. If the system’s capacity, in accordance with scenario B2 prediction, increases by 2050 to ~ 2000 GW(e), the annual uranium extraction would reach the level of 300−350 thousand tons per year, while the cumulative uranium consumption by 2050 would exceed 10 million tons. If the system’s capacity increases up to 5000 GW(e) by 2100, the annual uranium consumption would reach 750−900 thousand tons, and the integral uranium consumption would exceed 40 million tons, a value that is much higher than the assumption made above of available economically attractive resources. As it can be seen from Fig. 5.2 a resource potential of 16 million tons of natural uranium does not allow to keep the growth pace of thermal reactor capacity in accordance with the B2 scenario even until the middle of the century, since the lack of commercially attractive uranium fuel afterwards would have forced to phase out thermal reactors in a stepwise manner to provide nuclear fuel supply for a declining nuclear power system up to 2100. Commissioning after 2040 of fast reactors with a breading ratio 1.05 to be fueled with plutonium produced by their thermal predecessors and by themselves could in the best case only lead to a stabilization of nuclear power capacities at a level of 1500−2000 GW(e) (Fig. 5.2). Thus, under a reasonable assumption regarding the global commercially attractive uranium resource, the deployment of a nuclear system based on a business-as-usual approach in accordance with the SRES B2 scenario would not meet the INPRO criterion on availability of nuclear fuel for 100 years in the reviewed system. Measures of a different character have to be undertaken to fulfill the INPRO criterion. One principle direction would be to invest into geological exploration to increase assured commercially attractive uranium resources. Another direction would be to invest into development of a closed nuclear fuel cycle with fast reactors.
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As it has been shown in Chapter 3, for most countries of the Joint Study assurance of uranium supply for development of long-term nuclear programs is a challenging problem. The results of an analysis of the global situation for scenario B2 demonstrate that the problem cannot be solved at the expense of globalization of the uranium market, while a solution exists in changing the technological platform of nuclear power with the aim of radical increasing the efficiency of natural uranium utilization. Development of fast reactor technology provides one reliable option for technological innovations. Due to a fast breeder reactor’s physical advantages, low grade uranium ores unacceptable for use in thermal reactors could be used in a fast reactor system. That makes the power cost produced by this type of reactor practically insensitive to natural uranium market costs. There are enormous strategic resources of 238U to assure fuel supply for any nuclear programme in the world for a century and more. That’s why the decision to address development of an INS based on CNFC−FR by countries of the Joint Study seems absolutely reasonable from the standpoint of uranium resource assurance. However, realization of the strategy of efficient use of the resource of 238U in an INS based on CNFC−FR implies the implementation of timely measures for reaching long-term objectives. For example, Fig. 5.2 shows that after 2040 an introduction of fast reactors with a breeding ratio ~ 1.05 and three years of external cycle does not ensure implementation of the B2 scenario due to a shortage of plutonium extractable from spent fuel of light water reactors. It means that for a two-component system of thermal and fast reactors the problem of assurance of uranium fuel supply should be considered in a close link with the problem of assurance of plutonium fuel supply.
5.3.2. Assurance of plutonium fuel supply
As discussed in Chapter 3, scarcity of explored national uranium resources in the countries of the Joint Study with a relatively small operating nuclear power system but with high expectations of nuclear growth is one of the reasons for developing programs based on a low share of thermal reactors in their future nuclear power system. However, these countries have a small initial stock of fuel material fit for use in fast reactors, contrary to countries with a large fleet of thermal reactors established during a long time of development of nuclear power. This situation causes a problem of assurance of plutonium fuel supply for programs with a planned rapid deployment of fast reactors, which is especially acute for China and India. A technical solution of this problem consists in development of fast reactors with a high breeding ratio and short times of the external fuel cycle. However, there are other opportunities that could be very helpful in mitigating this problem. Predictions based on analyses of national strategies and scenarios indicate a fast growth of global accumulation of plutonium in spent fuel of thermal reactors. Hence, a comparison of national scenarios with global ones reveals a paradoxical situation where some countries in the world expect an acute shortage of fuels based on plutonium for their national energy programs while other ones are developing programs of disposal of excessive amounts of the same material. Therefore, contrary to the situation with uranium fuel, the problem of assurance of fuel based on plutonium is not a global one at least until the middle of the 21st century. Plutonium balances for national nuclear systems were studied during the first stage of the Joint Study with the use of the DESAE code [9]. It was shown that with a share of thermal and fast reactors in China’s nuclear programme given in Fig. 3.2 of Section 3.1.4 and under the assumption that by 2050 only fast reactors with a BR of ~1.05 and a three year external cycle would be operating the national nuclear programme would not be supplied with the
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necessary amount of plutonium for fast reactors fuel (Fig. 5.4, left). At the same time, with a small share of fast reactors in the world, a great amount of plutonium would be accumulated in the spent fuel of thermal reactors in case of realization of the global scenario B2 (Fig. 5.4, right).
FIG. 5.3. Plutonium balances for China and the world (left – China projections with a high nuclear growth, right – the world B2 scenario) Chinese specialists have demonstrated that under above conditions only a national scenario with about half the growth of nuclear power (Table 5.1) and a lower share of fast reactors than presented in Chapter 3 would be assured of nuclear fuel supply (Fig. 5.4). Table 5.1. The input data of the capacity of NPPs in the moderate scenario in China Year PWR，GW(th)/GW(e) FR，GW(th)/GW(e) Total, GW(e) 2000 6.3 / 2.1 0 / 0 2.1 2010 30 / 10 0 / 0 10 2020 120 / 40 1.5/0.6 40.6 2030 180 / 60 6/2.4 62.4 2040 240 / 80 60 / 24 104 2060 240 / 80 170 / 68 144 2080 240 / 80 320 / 128 208 2085 240 / 80 400 / 160 240 2100 240 / 80 400 / 160 240
FIG. 5.4. Plutonium balance for China in case of a moderate scenario.
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There are some opportunities to assure fuel supply in case of realization of scenarios with a higher rate of nuclear power deployment in China. One opportunity consists of reaching very high breeding characteristics in Chinese fast reactors within a short period of time. Another opportunity is a multilateral approach to the nuclear fuel cycle [33]. As it can be seen from Fig. 5.3, this approach under IAEA safeguard control potentially could provide a possibility to balance globally the production and consumption of plutonium and mitigate the problem of excessive production of plutonium in some countries and a deficit of fuel based on plutonium in other countries. Simultaneously, worldwide fuel services could become a part of a mechanism combining the technical characteristics of nuclear installations in different countries thus easing the problem of ensuring universal safety and fulfilling security requirements. In the whole, the experience gained in course of the implementation of the first stage of the Joint Study clearly indicated expediency to tailor national strategies to global perspectives.
5.4. GENERALIZED MODELS FOR GLOBAL ENERGY MIX SIMULATION
The first lessons of the Joint Study analysis of technical innovations in an INS CNFC−FR and their impact on the viability of the system identified some specific challenges resulting from the global character of the issues. One of the challenges is the evident need of elaboration of a generalized reference model of INS CNFC−FR to be used in regional and global scenarios. Indeed, it is not reasonable, and practically impossible, to try to include in medium and long- term regional and global projections characteristics of all national fast reactor designs and matching fuel cycles, especially taking into account significant uncertainties in technical and economic characteristics of innovative systems. As it was shown in previous chapters, a “step by step” approach induced to focus initially on the development of a joint vision of a near term INS CNFC−FR (the “lead INS CNFC−FR”). The data in Table 4.1 indicate that there are good preconditions for such an approach since generic directions for the first phase of INS CNFC−FR development are rather consistent. Parameters of the reference model of the lead INS CNFC−FR have to be based either on characteristics already reached in operating installations or on evolutionary designs. Evolutionary designs seem admissible if for achieving some of the indices of the reference model not too deep and expensive R&D would be needed to be carried out. Consensus regarding main characteristic of the INS CNFC−FR that is expected to follow the lead system would be very desirable for simulation of nuclear power at the time horizon beyond 2030−2040. Conceptual designs presented in Chapter 3 could form the base for this vision. As it was noted in the Chapter 4 reaching the consensus is a rather complicated task which requires more efforts. Rough generic indices of an INS can be agreed based on expert opinions in the course of further joint vision studies though a more clear picture will be received only based on deep, expensive and prolonged R&D. Considerations presented in the report are confined to a reference model of the INS CNFC−FR (Table 5.2) to be introduced during the first phase of serial fast reactors deployment (a near term INS CNFC−FR). Characteristics of the reference model were defined based either on operation experience, or on the data of evolutionary designs, or on expert expectations.
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5.5. REFERENCE MODEL OF INS CNFC−FR
5.5.1. Sodium cooled pool type reactor
The fast reactor will be of a pool type as more operating experience of prototype and demonstration size fast reactors are available for this design. The coolant will be sodium as sufficient maturity has been obtained with sodium, i.e. sodium technology has been mastered and sodium components have shown excellent performance in various reactors. Hence, the generic reactor will be of a pool type with sodium as primary and secondary coolant. Reactor power: minimum of two units, 1 GW(e) each To exploit the economy of scale, 1000 MW(e) is the desired option rather than going for a smaller size adopted for demonstration scale reactors. The operating experience from large size reactors has provided confidence in selecting 1000 MW(e) power output. Already many LWRs of 1000 MW(e) and more are in operation and size is not a constraint for their performance. Hence, 1 GW(e) is the desired size for a generic INS and there should be a number of units at the site with the minimum being two. TABLE 5.2. SPECIFICATION OF THE REFERENCE INS: CNFC−FR OF THE FIRST GENERATION
Reactor Parameter Reference data National variant Coolant Sodium Reactor configuration Pool Loop (Japan) Power plant With a minimum of 2 reactor units Thermal efficiency, % 43 > 39 (Rep. of Korea) Capacity factor, % 85 Life, a 60 Fuel MOX MOX in first development stage and metal in the next stage (China, India, Rep. of Korea) Construction time, 54 (from concrete pouring to first months criticality) Breeding ratio 1.2 1 (Rep. of Korea, France, Japan) Minor actinides None Yes (France; as an option: recycling Japan, Russian Federation, Rep. of Korea, China, India) Seismicity Parametric (depending on region) Burnup, GWd/t 150 (average) 120 (metal, Rep. of Korea) Specific steel 3.5 consumption, t/MW(e) Fuel cycle Fuel fabrication Pellet by powder metallurgy route Vibro compaction (Russian Federation) Injection casting (Rep. of
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Korea) Reprocessing Advanced aqueous process with Pyro (Russian Federation, Rep. Pu loss < 0.05 % and minor of Korea) actinide partitioning. U&Pu oxides to be co-precipitated in co-located and optimized reprocessing, re- fabrication and waste management facility for a number of reactors Cooling time before 4 1 (India, Rep. of Korea) reprocessing, years 2 for metal fuel and 4 for MOX (China) Solvent Tri-n-butyl phosphate or homologue U-Pu Separation Co-processing with no Pu separation Plant life, years 40 High level waste Glass vitrification management High Level Waste To be evolved (Volume/MW(e))
Thermal efficiency: 43 %
Increased plant efficiency decreases unit energy cost and reduces the thermal pollution levels of the environment. For demonstration fast breeder reactors, a steam temperature of around 766 K has been selected in order to restrict the hot pool sodium temperature to 820 K at nominal power. based on accumulated experiences on material data, revised high temperature design code rules and the international trend, it is possible to operate at a high hot pool temperature so as to generate steam at 803−813 K matching the steam conditions of fossil fuelled power plants, and leading to an increased plant efficiency in association with reduced thermal pollution. Hence, a plant efficiency of 43 % should be the target. Capacity factor: 85 % Over 300 years of operational experience with fast reactors worldwide justifies confidence that higher capacity factors could be obtained. Over the years, fast reactor technology has attained maturity and potential causes of problems have been identified and technological solutions have been evolved, for example for steam generator leaks. On the conventional side, thermal reactors have achieved high capacity factors. Operating experience of fast reactors has shown an excellent performance of components. Hence, capacity factors around 85 % should be the target. Life: 60 years Based on detailed studies on material data and structural mechanics considerations, 60 years design life seems possible to be demonstrated. Further, with a critical re-look on the number of design basis events (DBE) to reduce fatigue damage and with the choice of better materials with high creep strength, even a higher service life can be envisaged. Degradation mechanisms like corrosion, erosion, irradiation effects, etc., are issues in fast reactor systems, which need to be crucially evaluated, tested and assessed for 60 years.
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Fuel: MOX, metal as separate study module The choice of fuel will be oxide because oxide fuel technology has been well demonstrated and understood. Moreover, with a view to exploit the existing industrial infrastructure and capabilities and consequent economic considerations, oxide fuel is the best choice for fast reactor fuel. A major feature in favour of the choice of oxide fuel is its potential to go for a high burnup as its degradation mechanisms have been well modelled and through adoption of a suitable design, a very high burnup is possible to be achieved. Nevertheless, based on growth considerations especially for Asian countries, metal fuel is identified as an optional fuel. Construction time: 54 months until criticality Construction time could be definitely reduced using international experience. Though different countries have needed different times for construction due to a variety of reasons connected with industrial capabilities, etc., it has been shown that reactors with 500−1000 MW(e) can be constructed in about 5−5.5 years. With the advent of modern construction technology and efficient project management techniques, 54 months is considered to be an achievable parameter. Parameter for breeding ratio or actinide burning It is well known that oxide fuel has many technological advantages and rich operating experience and hence was chosen for demonstration reactors towards techno-economic feasibility without emphasis on breeding. For the proposed INS, use of oxide fuel is continued as there is a scope to reduce doubling time by suitably optimizing various fuel designs and fuel cycle parameters, with a target doubling time of 40 years. For oxide fuels, the breeding ratio should be around 1.2. The reactor would be designed in a way that it can work with metallic fuels also after replacing the oxide core. R&D is to be pursued to select optimized parameters for operation of metallic fuelled reactors. The optimization would cover parameters such as composition, fabrication methodology, fuel design options, burnup and breeding ratio. The breeding ratio targeted would be around 1.6. Considering options regarding the fuel, the breeding ratio could be a variable parameter. Suitable parameters should be developed for an INS based on fast reactors burning actinides and other long lived fission products. Seismicity The seismicity should be a variable parameter as it depends on the site of installation in a given region. Hence, it should be a variable parameter in the study. Innovative engineering design features Major features incorporated to achieve economic competiveness are a compact main vessel without cooling, a thick plate concept for top shield, an anchored safety vessel, a longer plant life, a compact secondary sodium piping layout with a minimum number of steam generator modules, in-vessel primary sodium purification, compact plant layout, use of advanced structural materials for clad and wrapper, minimizing the use of SS316 LN material for nuclear steam supply system components and concrete for civil structures, adoption of a rationalized state-of-art seismic design approach and a twin unit concept. Burnup: (parametric: 150 and 200 GWd/t average) Enhancing the target burn-up of the fuel has a direct effect of reducing unit energy cost. In view of the large experience in the world, especially in the USA, the Russian Federation and France, gained with respect to improved cladding and wrapper materials, a higher burn up to
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150 GWd/t would be considered in the initial phase with an ultimate target of 200 GWd/t. Choice of advanced materials and improvement of present materials through characterization, design of materials, modelling and performance studies constitute the primary mechanism of achieving economy through a higher burnup.
5.5.2. Fuel cycle plants
The reference MOX based 1 GW(e) fast breeder reactor with a closed fuel cycle is assumed to be co-located with fuel fabrication and reprocessing plants of matching capacity in a single physical protection boundary. Therefore time delays, associated with multiple protection boundaries, are avoided. The generic INS in this case consists of three integrated facilities: a fuel fabrication plant, a reprocessing plant and waste management facilities. After the first few years of operation, the fast breeder reactor with a closed fuel cycle is assumed to be self- sustaining for inventories of Pu and the only input required will be a make-up of uranium to compensate for consumption in fission and breeding processes. Capacity for fuel fabrication and reprocessing plants The capacity of the fuel fabrication and reprocessing plant is much lower in case of the reference INS as compared to those needed for a system with conventional water-cooled reactors because of the very high burn-up achieved in fast breeder reactors. In case of a fast reactor park, consisting of co-located identical reactors, the design of fuel fabrication and reprocessing plants would be based on a modular concept for cost benefit. The reference INS fuel-cycle facility would have a capacity to meet the requirements of 4 reactors of 1 GW(e) each, e.g., the quantity of discharged fuel from each 1 GW(e) reactor is assumed to be 5 tons of heavy metal per year. Thus for the fast reactor park consisting of 4 reactors of 1 GW(e), the fuel cycle facility should have a processing capacity of 20 tons of heavy metal per year. Optimizations for the plant sizing would be primarily done for the reprocessing plant because due to a higher burn-up of the fuel being processed and subsequent higher activity handled, it requires intensive shielding. These optimizations of process technology including the equipment for solvent extraction operations could be performed on the following basis: on waste volume produced per ton of heavy metal processed and on volume of shielded cells per ton of heavy metal processed. In this context, it may be mentioned that the concurrent solvent extraction technology (using centrifugal extractors) does not require as huge a headspace or as large a footprint as the earlier generation equipment (pulse-columns and mixer-settlers) required. However a bottleneck arises from the criticality point of view as for plutonium containing solutions, slab tanks or annular tanks with a high surface to volume ratio are needed. Use of borated/poisoned steel as structural material for tanks as well as tanks with poison tubes will result in safe higher holding capacities. Fuel fabrication The fuel fabrication for the fast reactors would be based on a mixed powder route. Mixed oxide could be made by co-processing and co-precipitation and this mixed oxide product may be suitably diluted by adding UO2 powder to produce fuel for multiple compositions of fast reactor cores. Since U–Pu separation is not envisaged, several process steps are eliminated resulting in a reduced number of process equipment, tankage and operations, leading to a significant reduction in processing costs.
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Reprocessing of spent nuclear fuel In the present context, the reprocessing operation involves recovery of unused as well as bred fissile materials as well as recovery of minor actinides and selected long-lived fission products in a form suitable for immediate recycle in the reactor or co-located transmutation systems. It is assumed that aqueous processes can be used for a tentative burn-up of 200 GWd/t and a 360 days cooling period for discharged fuel.
Solvent extraction process for reprocessing
The solvent extraction processes to be used in the reference INS will be an advanced PUREX process. The advanced PUREX process is a superset of the classical PUREX process with several novel features envisaged. Its constituent radical approaches aim at a cost effective combination of waste treatment processes as well as step changes in process and waste technology to achieve significant cost savings. The main differences from the conventional PUREX process are: • High efficiency contactors (centrifugal contactors) will be used in the solvent extraction operation to limit radiation-induced solvent/diluent degradation and Zr extraction by having a small residence time. Smaller holdup and lower inventory are additional benefits. Use of centrifugal extractors will also enable highly economical civil construction for the reprocessing facility as centrifugal extractors do not require a tall headspace and height leading to a drastic reduction of the volume of the shielded radioactive cell; • Based on the observed performance of centrifugal contactors, it may be assumed that use of only two cycles of centrifugal extractors based primary solvent extraction (with TBP) will result in a product with designated purity and a sufficient decontamination factor for safe handling in the adjoining fuel fabrication glove-box trains; • As stated earlier, U-Pu separation is not envisaged resulting in economic benefits; • In the advanced PUREX process, the primary on TBP or its homologue based extraction could result in the additional separation of neptunium, tritium and technetium (99Tc) in the high active waste (HAW) cycle itself. Tritium and technetium may be reclaimed from the aqueous nitric solutions by additional chemical processes. By confining the tritium to the high active waste cycle only, the plant wide contamination by tritium is avoided. Technetium will be converted to metal form and has to be irradiated at peripheral positions (blankets) of fast reactor cores for transmutation. Another long lived fission product 129I will be absorbed in caustic solutions and the resulting NaI has to be irradiated also in transmutation systems; • The major emphasis is given on a simplification of the process flow sheet, reduction of the number of process cycles as well as process steps and effective cost cutting due to implementation of a simpler technology. It may be mentioned that in the aqueous route of reprocessing, an extremely high separation factor (also called decontamination factor) of 107 and a high recovery rate ≥ 99.8 % are routinely achieved. For the reference INS stipulated Pu recovery rate is 99.95 % or more. Emphasis is put on robust online analytical techniques for monitoring of Pu in raffinate streams and activity levels in the loaded organic streams and subsequent fine-tuning of operating parameters to achieve a significant reduction in the rework of off-grade products. It would also result in increased plant throughput and availability.
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Separation of minor actinides from waste shall be integrated in the proposed reprocessing not only for transmutation but also for reduction of the storage volume of wastes that are to be monitored for a very long time. Recently, several new extractants for such a separation process have been reported. To achieve an actinide-free status for high active waste, recovery levels of minor actinides are assumed to be 99.9 %. Waste management The current paradigm for waste management plants is mainly based on the following points: • Safety of operators as well as general public by protection against chemical as well as radiological toxins; • Design should yield a minimum of waste management costs for the lifetime of the plant; • The plant should release minimum effluents; • Active process waste must be conditioned to meet safe interim storage and transport requirements. In addition, it has to fulfil anticipated final disposal requirements for long time. These requirements have necessitated the separation of actinides and long lived isotopes. All waste streams are managed so as to reduce waste at the source. In addition, almost all are treated up to the point of discharge or up to the production of an encapsulated solid waste form suitable for disposal. The main goal of solvent extraction based minor actinide separation is: • Minimization of the long-term radiotoxic inventory of wastes conditioned in conventional matrices like glass. • Minimization of the heat load of conditioned wastes (requires removal of 137Cs and 90Sr) to reduce cooling requirements of the repository. For the reference INS, solid waste volumes shall be equal or less than accepted benchmarks. In addition additional waste reduction features like an acid recovery/recycle and online solvent purification by improved methods like vacuum distillation would significantly reduce the waste volumes produced. Materials of construction, plant life and its extension As the prime chemical, used in the aqueous route of reprocessing, nitric acid is quite corrosive at the range of concentrations handled; thus, materials of construction used in the fabrication of the plants should be resistant to corrosion. The problem of corrosion resistance is made more complex by acute effects of radiation and organic vapours, especially Tri-butyl phosphate (TBP). For metallic parts, corrosion could be minimized using the NAG (nitric acid grade) steels, e.g., austenitic SS304L NAG. On the other hand, radiation and TBP vapours attack almost all polymers used in cable insulations, flexible seals, soft tubing and electronic devices. Therefore use of polymers is kept to the minimum possible and related items are designed suitable for remote maintenance. In some cases, special materials for construction with an extremely low corrosion rate for crucial equipment have been tried, e.g., titanium-tantalum-niobium alloy and zirconium for dissolvers where spent nuclear fuel is in contact with 8-12 N nitric acid at elevated temperatures. Unlike normal chemical plants, reprocessing plants have some unique characteristics, like a several kilometers long tubing of narrow diameter inside a shielded-cell area, indirect ways of liquid transfer by airlifts, and vacuum assisted and fluidic devices with sufficient redundancy. Some of the paraphernalia have a service life identical to the plant life wherever some parts of
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plants are designed for remote maintenance and replacement. The reprocessing plant (a component of the INS) may have a design life of 40 years, comparable to the reactor lifetime. A lifetime extension is possible by periodic condition monitoring of the plant constituents by having in-service inspection by remotely controlled devices like robot mounted cameras, special fiber-optic probes for visual inspection and ultrasonic measurements for wall thinning. In a waste management plant, the main problem of corrosion is due to molten glass. In this context, relevant parts of the plant are to be designed for remote maintenance and replacement. Advanced technologies like the cold-crucible technique would reduce the effect of corrosion and ensure a long life with economic benefits. Separation of heat emitters and glass-incompatible materials from fission products would ensure a higher loading in the glass matrix and would result in a significant reduction of the final conditioned waste volume. Construction time A construction time of 72 months is assumed to be achievable from the first pouring of concrete to active commissioning. Modern project scheduling, management and erection methods may reduce it to 60 months as a minimum. It is essential to complete construction and hot commissioning of the fuel cycle facility along with the reference reactor so that discharged fuel could be processed immediately after a short cooling period and out of pile inventory of fissile material could be kept to a technically achievable minimum.
5.5.3. Experience and knowledge
The ultimate goal of the application of the INPRO method is to confirm that the INS assessed fulfills all the INPRO criteria and therefore represents a sustainable system for a Member State (or group of MS) [2]. Reliability of input data used in an evaluation determines reliability and value of the evaluation. An INS based on fast reactors with sodium coolant can use technical, economic and environmental data supported by real industrial experience of system operation in different countries of the world. There are commonalities and diversities in approaches to develop such a system as indicated by Member States. It is also clear that a large knowledge base exists. As an example, Fig. 5.5 shows how knowledge was accumulated both by “learning by doing” and “learning by searching” and how it enabled bringing projected specific capital costs of sodium cooled fast reactors) closer to the level of water cooled thermal reactors.
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Evolution of SFR/WTR spesific cost 3.5
3
SPX 2.5
2
1.5
Relative units BN-600
EFR BN-800 PFBR 1 BN-1800 JSFR SFR/WTR=1 0.5
0 1965 1970 1975 1980 1985 1990 1995 2000 2005 Project year
FIG. 5.5. Evolution of specific capital cost of sodium cooled fast reactors (SFR) with respect to water cooled (WTR) thermal reactors. (BN = Russian fast reactors; EFR = European fast reactor; JSFR = Japanese sodium cooled fast reactor; PFBR = prototype fast breeder reactor in India; SPX=SUPERPHENIX in France) Accumulation and exchange of experience and knowledge of fuel cycle technology is also very valuable. Fuel cycle costs have an important influence on the cost of generated electricity and could amount up to 20−25 % of the latter. Fuel cycle costs consist of the costs for fuel fabrication, reprocessing operations and radwaste management. It is worth noting that for the same quantum of generated electrical energy, fuel fabrication, reprocessing and radwaste management costs for a fast reactor fuel cycle are a fraction of similar costs for a conventional water-cooled reactor fuel cycle. The prime reason for this cost reduction is low material requirements due to the ultra-high burn-up achieved in a fast reactor fuel cycle. Further cost cutting could be achieved through the following means: • By simplification of technology like a reduction of the number of process cycles/process steps and implementation of innovative flow-sheets with tighter control over operational parameters; • By achieving a higher level of automation in fuel fabrication, post irradiation and waste package handling; • By achieving an optimum level of human resources and knowledge management. By suitable conversion of operating experience into a retrievable electronic knowledge form, individual achievements, experiences and subjective observations leading to successful diagnostics of process upsets, equipment malfunctions and other operational bottleneck could be stored for future reference and this valuable information would not be lost when persons retire or move on. It could avoid duplication of diagnostic efforts and rediscoveries leading to a significant reduction of downtime of the plant. The global nuclear community has the capacity to leapfrog making an INS based on CNFC−FR cost competitive and capable to meet all other requirements of sustainable energy supply. An assessment of this INS − that will be implemented at the second stage of the Joint
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Study − has to recognize all national R&D directions, which can become a basis for collaboration developing a robust and competitive INS CNFC−FR. The advantages of an INS unification along with the complementarity of national conditions form a great reserve for accelerating the deployment of the INS. Joint considerations and assessment of the INS can become a first step on the way to a joint development of generic designs and technologies, and eventually lead to a joint fabrication of unified equipment and fuels.
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CHAPTER 6 ECONOMICS
6.1. INTRODUCTION
This chapter outlines results obtained by participants of the Joint Study by an assessment of economics of INS CNFC−FR using the INPRO Methodology [2, 5, 35, 41]. The judgement procedure for assessing the capability of an INS to comply with the INPRO hierarchy of demands starts with the evaluation of the INPRO criteria (bottoms up approach). It is assumed that, if all criteria are fulfilled, the corresponding user requirements are fulfilled, as well as the basic principles [35]. In accordance with guidance for the application of the methodology, consideration of some context issues has to forestall the economic assessment itself to provide general background and boundary conditions important for the evaluation. In Section 6.2 general considerations of the Joint Study participants are briefly addressed. Section 6.3 describes the main context of the study by presenting macro-economic and other boundary conditions, including existing national power systems, electricity markets, nuclear power status and structure, which needs to be taken into account when carrying out an economic assessment. Description of the main parameters of the generalized reference model of the INS CNTC−FR related to economic assessment is briefly presented in Section 6.4, along with comments on important deviations from the model considered by some countries participating in the Joint Study. Section 6.5 presents results of the evaluation of the INS economic indicators values obtained by the Joint Study participants. In addition to the results of the economics assessment, findings and considerations of the Joint Study participants on further improvement of INS CNFC−FR with regard to economics are given in Section 6.6, which also includes a feedback to INPRO methodology improvement in the economic domain.
6.2. CONSIDERATION OF THE ASSESSORS
Before proceeding with a description of input data for the economics assessment of the INS CNFC−FR using the INPRO Methodology, it is necessary to present general considerations of the assessors – the countries of the Joint Study. It should be noted that all the Joint Study countries possess large-scale electric power systems, with some (dominant, in case of France) share of nuclear power. The majority of the Joint Study countries have implemented extensive programs aimed at further deployment of nuclear reactors and the realization of a closed nuclear fuel cycle. Finally, they represent practically the entire international community with a concrete interest in development of fast neutron liquid metal cooled reactors. At the same time, Joint Study national teams involved in carrying out economics assessment of INS CNFC−FR were in an in-between category with regard to the two basic types of assessors identified in the INPRO Manual [35]: type 1 – an assessor who is looking at deploying an INS, and type 2 – an assessor who is determining whether to invest in developing an INS. The need and expediency of further developing and deploying both fast reactors and closed fuel cycle technologies is recognized in the majority of the countries participating in the Joint Study. However, it is not an indispensable choice for the countries of the study that do not develop the technology. Different angles of view enabled finding of balanced assessments and judgments.
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6.3. INPUT FOR ASSESSMENT: MACRO-ECONOMIC AND BOUNDARY CONDITIONS
According to the INPRO manual on economics [41], for an assessor who intends to apply the INPRO methodology to confirm that an INS complies with his expectations and goals, it is of high importance to create, to the extent possible, a complete clear vision of the power system in which an INS is supposed to be deployed. The components of such a vision include, but are not limited to, the structure of the power system, availability and affordability of fuels and energy resources, nuclear power reactor and fuel cycle status and options, projected energy demand and supply, nuclear power deployment scenarios, etc. The principal purpose of such “context” considerations is to help an assessor to better formulate desirable targets and recognize possible limitations thus enabling a more realistic judgment on the potential of an INS to meet an assessor’s specific requirements. Comprehensive context data on macroeconomic and boundary conditions for each country of the Joint Study looking at deploying the INS CNFC−FR have been presented in Chapter 3 of the Report. With reference to the data and information taken from Refs [19], [37], [38], [39], and [40] a brief survey of commonalities and differences in the energy sector of the countries participating in the Joint Study is done below in Subsections 6.3.1 to 6.3.4.
6.3.1. Power systems and energy resources
As it was mentioned in Section 6.2, a common feature of the countries participating in the Joint Study is that each of them has a large-scale power system. However, the difference in power system structures among the Joint Study countries is significant enough to result in different national strategies and approaches to nuclear power deployment and achievement of long-term goals. One of the principal objectives of Joint Study countries is to ensure national energy security and power system sustainability for a long-term perspective. Practically, this could be achieved through a reduction of dependence of national power systems from importing and prevailing use of natural fossil energy sources (coal, natural gas, oil) and a balancing of power structure. This consideration is illustrated by Fig. 6.1, which demonstrates comparison of primary energy consumption patterns in the Joint Study countries [19].
100% 90% 80% 70% 60% 50% 40% 30% 20% 10% 0% China France India Japan Korea Russia Canada Ukraine Oil Natural Gas Coal Nuclear Energy Hydro electric FIG. 6.1. Comparison of primary energy consumption patterns in Joint Study countries.
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The share of fossil-fired power in most cases (except for Canada and France) exceeds 80 %.
The amount of domestic energy resources available within the Joint Study countries differs significantly – from huge to negligible (see Table 6.1). In any case, the energy policy of the countries is focused on extension of the resource base. Alternative energy sources like wind, geothermal, solar and biomass are recognized to provide good opportunities as a local small- to-medium size power source, but they are not capable of meeting the increasing energy demand in a comprehensive manner.
TABLE 6.1. PROVEN RESERVES OF MAIN PRIMARY ENERGY RESOURCES IN THE COUNTRIES OF THE JOINT STUDY (2006) [19], [37]
Hydropower, Oil, Natural Gas, Coal, terawatt-hours/year thousand trillion million tons Gross theoretical Technically million cubic capability exploitable barrels meters capability China 16.3 2.45 114 500 6 083 2 474 France 0 0 15 270 100 India 5.7 1.08 9 2445 2 638 660 Japan 0 0 359 718 136 ROK 0 0 80 52 26 RF 79.5 47.65 157 010 2 295 1 670 Canada 17.1 1.67 6 578 2 216 981 Ukraine 0 1.10 34 153 45 24 ROK = Republic of Korea, RF = Russian Federation. Addressing a nuclear source seems inevitable for the countries having large-scale power systems and a lack of indigenous energy resources. However, even for the countries possessing substantial reserves of fossil fuels and at the same time having large-scale national power systems, the use of nuclear power is advantageous since this option provides an opportunity to enhance energy security and sustainability features of the domestic energy sector while preserving its export potential. The need to scrutinize the potential of nuclear power technologies as an alternative to traditional ones becomes more and more important in this situation.
6.3.2. Electricity markets
An assessor needs to have a general understanding of the electricity market conditions in the country where an INS is expected to be deployed. This knowledge would be useful for an initial assessment of business prospects, including long-term perspectives, as well as for various risk analyses. From the viewpoint of the INPRO methodology, such aspects are reflected in the user requirements UR2 (ability to finance), UR3 (investment risk) and UR4 (flexibility). By generalizing the electricity market information of the Joint Study countries, one could conclude that in many countries a process of electricity market reform goes on. This reform is being carried out by taking into account different experience gained all over the world which demonstrates both evident achievements and quite serious problems on the way of building a competitive market for electricity supply.
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Indeed, the new conditions of de-regulated markets of generation, transmission and distribution of electricity facilitate involvement of private players and expansion of opportunities for private-government partnerships. Strong competition in those markets is expected to result finally in reduction of prices for consumers. Besides, a competitive market is obviously advantageous for a consumer since it allows planning of power supply more efficiently and it is a more transparent system with a possibility of choice. At the same time, electricity is a very specific kind of goods, and consumers of electricity might be concerned not only with the price but also with such “trumps” of regulated power systems like reliability, stability and security of electricity supply. Some information concerning specific features of today’s electricity system in some countries participating in the Joint Study is provided below. China
In 2002, the Chinese government dismantled the monopoly state power corporation (SPC) into separate generation, transmission, and services units. Since this reform, China’s electricity generation sector is dominated by five state-owned holding companies, namely China Huaneng Group, China Datang Group, China Huandian, Guodian Power, and China Power Investment. These five holding companies manage more than 80 percent of China’s generating capacity. Much of the remainder is operated by independent power producers (IPPs), often in partnership with privately-listed arms of these state-owned companies. Deregulation and other reforms have opened the electricity sector to foreign investment, although this has so far been limited. During the 2002 reforms, SPC divested all of its electricity transmission and distribution assets into two new companies, the Southern power company and the State Power Grid company. The government aims to merge SPC’s 12 regional grids into three large power grid networks, namely a northern and northwestern grid operated by State Power Grid company and a southern grid operated by a Southern Power company. Chinese officials hope to achieve an integrated national electricity grid by 2020. Also in 2002, the State Electricity Regulatory Commission (SERC) was established, which is responsible for the overall regulation of the electricity sector. France
France has the second-largest electricity sector in the EU, after Germany. In 2004, France produced 540.6 billion kWh of electricity and consumed 440.6 billion kWh. The country depends upon nuclear energy for 79 percent of its electricity generation. Electricite de France (EdF), owned by the French government, controls almost the entire market for electricity generation and distribution in the country. Gestionnaire du Reseau de Transport d'Electricite (RTE), a company nominally separate from but controlled by EdF, operates the national electricity grid. EdF has a monopoly on electricity generation and distribution in France. Facing criticism from the European Commission, France has begun to slowly privatize EdF and to open the electricity sector to other companies. Compagnie Nationale du Rhone (CNR) is the second- largest electricity utility in France, operating 19 hydroelectric plants on the Rhone River. Deregulation has also allowed the entry of small distributors into the market. These distributors purchase wholesale electricity from EdF, and then sell it to large industrial and commercial customers.
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Japan Japan’s electricity industry is dominated by 10 privately-owned, integrated power companies that act as regional monopolies, the largest of which is the Tokyo Electric Power Company (TEPCO). These companies account for more than three-quarters of Japan’s electricity capacity and also control the country’s regional transmission and distribution infrastructure, leaving limited room for independent power producers (IPPs). The 10 companies cooperate with one another to ensure the stability of electricity supply and work together to exchange or provide electricity during emergency situations or power shortages. Other players in the electricity market are the Japan Atomic Power Company (JAPC), which operates three nuclear power plants, and the Electric Power Development Company (known as J-Power), which operates 16 GW of hydroelectric and thermal power plants. J-Power, formerly a state- owned enterprise, was privatized in September 2004.
While Japan’s 10 regional power companies are privately owned and subject to some competition, historical regulation of the power sector has guaranteed effective monopolies for the companies. Japan has set out to liberalize and deregulate the electric power sector on a step-by-step basis. The Japanese Diet passed a bill in May 1999 that amended the Electric Utilities Industry Law to allow a partial opening to competition. Beginning in March 2000, large industrial power purchasers, representing about one-third of the Japanese electricity market, have been able to choose their power suppliers. In April 2005, the scope of liberalization was increased to include medium-scale electricity purchasers. The Japanese government intends to consider full liberalization beginning in April 2007. In February 2004, the Electric Power System Council of Japan (ESCJ) was established to oversee the liberalization and deregulation program. The Ministry of Economy, Trade and Industry (METI) has primary regulatory authority for the energy industry. Within METI, the Agency of Natural Resources and Energy (ANRE) and its various electric power subdivisions oversee the electricity sector. Japan’s Nuclear Safety Commission regulates operations at the country’s nuclear power plants. Republic of Korea
Republic of Korea’s government intends to privatize the country’s power sector, although these plans have proceeded slowly due in part to tepid investor response, public concerns about rising power prices, and reluctance on the part of the government to sell some power sector assets. In 2001, state-owned Korean Electric Power Generating Corporation (KEPCO) was broken up into six separate power generation companies (“gencos”) in preparation for privatization. Of these six, Republic of Korea’s government will retain control of one “genco”, the Korean hydro and nuclear power company (KHNP), which operates the country’s nuclear and hydroelectric power stations. The remaining five are to be auctioned off, although the planned sales have been repeatedly delayed since 2002. Current plans call for the government to sell majority stakes in the “gencos” to private companies and offer minority stakes through public stock issues. Foreign ownership is limited to 30 percent of any single division. The bidding process will be managed by the Korean Power Exchange (KPX), which was established in 2001. All told, KHNPC owns four nuclear power plants totaling 20 nuclear reactors, with three additional units currently planned or under construction. Investment in new conventional thermal power plants has subsided in recent years as a result of uncertainty surrounding the planned privatization process.
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Russian Federation There are seven separate regional power systems in the Russian electricity sector: Northwest, Center, Middle Volga, North Caucasus, Urals, Siberia, and Far East. The Far East region is the only one not connected to an integrated power system. The country's largest generation owner, called the Unified Energy System of the Russian Federation (UES), is 52 percent owned by the Russian government and controls most of the transmission and distribution in the Russian Federation. UES owns 96 percent of the transmission and distribution system, the central dispatch unit, and the federal wholesale electricity market (FOREM). The Russian grid comprises almost 2 million miles of power lines, 93 000 miles of which are high-voltage cables over 220 kilovolts. As part of the reform begun in March 2004, Russian President Vladimir Putin signed six bills into law that aim to substantially reform the industry. Under the new laws, tariff rates on the domestic market are to be made more universal instead of geographically-specific. In November 2006, Energy Minister Khristenko said that electricity rates will increase by 10 percent in 2007 and that around 5 percent of the electricity market will be liberalized beginning January 1st, 2007. The reform also calls for UES's generation and distribution facilities to be privatized, but the country's transmission grid will remain under state control. The main goal of the Russian electricity reform package is to create a generating sector divided into multiple wholesale electricity companies (commonly called "OGKs"), which participate in a new competitive wholesale market. In September 2006, the Russian cabinet finally approved the spin-off and privatization of two generating companies (gencos), Wholesale Generating Company (OGK)-5 and Territorial Generating Company (TGK)-5, as well as additional 11 gencos to be spun off during 2007. The UES reform plan will distribute UES stakes in 20 gencos on a pro-rata basis to current UES shareholders. The State, with 53 percent ownership, is the largest shareholder in the UES. The current plan is to transfer the State’s share in the gencos to two companies, the Federal Grid Company and the Hydro-OGK, which will remain state-controlled after UES ceases to exist on July 1st, 2008. The goal is for the market to become completely liberalized by 2011.
Ukraine Ukraine's power sector is the twelfth-largest in the world in terms of installed capacity, with 54 GW(e). The country is currently in the process of revamping its electricity sector, through privatization, increased utilization of existing facilities, and the completion of two new nuclear plants. In Ukraine, thermal power plants (oil, natural gas, coal) account for nearly 50 percent of power generation, with nuclear power generating another 40 percent, and hydroelectric generation accounting for approximately 10 percent. Ukraine has four operating nuclear power plants. These plants have a combined capacity of 12.8 GW(e), accounting for approximately 24 percent of the country's total power-generating capacity. Ukraine's nuclear plants produced 40 percent of the country's power in 2004. Currently, only six Ukrainian distribution companies have been fully privatized, and 20-45 percent stakes in nine other utilities were sold in 1997−1998. Further privatization of the sector is currently not planned. Since 1997, the Ukrainian National Electricity Regulatory Commission (NERC) has facilitated a centralized market for wholesale electricity, called the Wholesale Electricity Market (WEM). Power producers sell into a common market, operated by Ukraine's wholesale electricity operator, Energorynok, and distribution companies distribute the power to the end user. Although the government fixes the price of nuclear and
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hydro generation supply, the market has made progress in basing wholesale electricity prices on an upcoming hour's and an upcoming day's basis by using bidding prices of electricity production. A tariff methodology was introduced in 2001 for privatized utilities concurrent with the sale of controlling stakes in six utilities. Conclusion To sum up the survey, it should be concluded that eventually the trends in the energy and electricity markets may influence approaches designing nuclear systems. Indeed, some experts even supposed that a de-regulated market could request nuclear units of small and medium capacity which would require smaller investments for their construction, thus facilitating access for private energy companies into nuclear business. Nevertheless, participants of the Joint Study agreed that nuclear units of 1 GW(e) and more will remain the backbone of nuclear power for a long time and most electricity markets will adopt these units.
6.3.3. Nuclear power, reactor types, fuel cycle schemes
Existing nuclear power systems play an important role in each of, though the share of nuclear electricity generation ranges significantly, from about 2 % in China up to nearly 80 % in France. Thermal (boiling and pressurized) light-water reactors of various designs dominate in the Joint Study countries, operating mainly in a base load mode with an once-through fuel cycle and using uranium oxide fuel, except for Canada and India where pressurized heavy water reactors are exploited. A summary of the reactors in operation in the Joint Study countries is shown in Table 6.2. While using nuclear power in an open fuel cycle, a majority of the Joint Study countries keep conducting R&D in the field of nuclear reactor and fuel cycle technologies aimed at a transition to a closed fuel cycle. Fast reactor technology has been demonstrated through more than 50 years of operation of experimental and prototype reactor facilities in France, India, Japan and the Russian Federation. China with the CEFR reactor under completion is also on the threshold of entering the fast reactor community. TABLE 6.2. GROSS CAPACITIES OF NUCLEAR POWER REACTORS IN OPERATION IN THE COUNTRIES OF THE JOINT STUDY, MW(e) (2007) [38]
ABWR BWR FBR PHWR PHWR/ PWR PWR/ PWR/ LWGR/EGP LWGR/RBMK CANDU VVER VVER 1000 440 China, - - - - 1400 54682000 - - - mainland Canada - - - - 13817 - - - - - France - - 250 - - 66056 - - - - India - 420 - 3995 ------Japan 6550 23664 - - - 19366 - - - - ROK - - - 2785 - 15339 - - - - Russian - - 600 - - - 9000 2618 48 11000 Federation Ukraine ------13000 880 - - ROK = Republic of Korea. Achievements in the field of fuel cycle technologies are also remarkable. France implemented at an industrial scale (~1700 t(HM)/a) reprocessing of irradiated nuclear fuel from domestic and foreign thermal reactors with subsequent use of regenerated uranium and plutonium in the form of mixed oxide fuel again in thermal reactors for a limited number of recycling (1
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times), thus practically approaching a closed scheme of a nuclear fuel cycle. In November 2006 Japan reported that Rokkasho nuclear fuel reprocessing plant having a design capacity of 800 t(HM)/a yielded its first production of mixed uranium-plutonium oxide powder. The plant does not produce any separated (i.e., pure) plutonium, as it uses the U-Pu co-denitration process, a proliferation-resistant technology developed in Japan. The Russian Federation operates a reprocessing plant with a design capacity of 400 t(HM)/a at Mayak, Chelyabinsk region, primarily for reprocessing spent uranium oxide fuel from VVER-440 type light water reactors designed in the USSR, and uses regenerated uranium to produce fresh fuel for RBMK (graphite moderated) reactors. India has several reprocessing plants, including one at Kalpakkam, having a design capacity of 275 t(HM)/a.
6.3.4. National deployment scenarios
National scenarios of possible nuclear power deployment based on INS CNFC−FR are presented and discussed in detail in Chapter 3 of the report. Analysis of these scenarios allows to conclude that regardless of the nature of changes envisioned in national nuclear power systems either for substitution of existing reactors or for commissioning of new capacities, each of the Joint Study countries is expected to face considerable capital investments in construction of power reactors. In addition to that, most of the Joint Study countries will need to create fuel cycle capacities matching the anticipated reactor park extension.
6.4. INS DESCRIPTION
6.4.1. Reference INS CNFC−FR
Approach to and results of the development of a reference model of the INS CNFC−FR are presented in Section 5.5 of the Report. An important feature of the reference model is that it defines an INS parameter range in which all national systems are included. Table 6.3 enlists some parameters of the reference model required for carrying out an assessment using the INPRO methodology.
TABLE 6.3. PARAMETERS OF THE REFERENCE INS CNFC−FR RELATED TO THE ECONOMIC AREA
Parameter Value Reactor Power Minimum 2 units, 1−1.5 GW(e) each Thermal efficiency, % 43 Capacity Factor, % 90 Life, years 60 Overnight Capital Cost per Benchmark with reference water reactor in each Joint kW(e) Study country Construction time, months 54 (from concrete pour to first criticality)
6.4.2. Peculiarities of national INSs
A few important deviations from generic core features of the INS under consideration were determined for further assessment. Some of these peculiarities being important for performing an economics assessment are presented in Table 6.4.
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TABLE 6.4. DEVIATIONS OF PARAMETERS OF NATIONAL INSS FROM THE PARAMETERS OF THE REFERENCE MODEL OF THE INS CNFC−FR
Reference INS France India Japan ROK RF Reactor power, MW(e) 1000−1500 1500 500 1500 600 1800 Time of construction, years 4.5 4 5 7 2.7 6 Discount rate, % - 8 12 2 8 5 ROK = Republic of Korea, RF = Russian federation. Deviations in some characteristics of INS components reflect preferences of Joint Study countries in applying the INS CNFC−FR thus stressing its flexibility and high potential to play a significant role in ensuring energy security and to help solving inherited issues from existing nuclear power systems related to spent fuel and radwaste management, as well as resource restrictions.
6.5. RESULTS OF INS CNFC−FR ECONOMICS ASSESSMENT
This section generalizes the results of an economics assessment of INS CNFC−FR obtained by national teams. The evaluated indicators from different national assessment studies are aggregated, analyzed and discussed. An example of a complete economics assessment of a national INS CNFC−FR comprised of tables containing judgments on all evaluated INPRO indicators and corresponding acceptance limits (AL) along with rationales can be found in Annex A and B In the area of economics, an INPRO assessor is required to determine the following [41]: • Whether the cost of energy, i.e. electricity, is cost competitive with the cost of electricity produced by the cheapest alternative means of generating electricity in the country, i.e., the cheapest alternative generating source (CAGS), (INPRO user requirement UR1); • Whether the values for some financial figures of merit used to judge investment payback for investment in nuclear power are comparable to the values for investment in CAGS, (UR2, criterion CR2.1); • Whether the total investment required to deploy a nuclear power plant can be raised in a given market, (UR2, criterion CR2.2); and • Whether the risk associated with the investment needed deploying the nuclear power plant is acceptable, (UR3).
6.5.1. Cost competitiveness
The definition of INPRO user requirement UR1 is [41]: The cost of energy from innovative nuclear energy systems, taking all relevant costs and credits into account, CINS, must be competitive with that of alternative energy sources, CA, that are available for a given application in the same time frame and geographic region:
CINS < k * CA , where k is a factor reflecting preferences of Member States or investors depending on their particular circumstances. The factor k can be one, less than one, or greater than one in a given Member State or region, depending on whether or not nuclear costs are offset by credits relative to an alternative energy source or vice versa. Reaching almost cost competitiveness of the CNFC−SFR is demonstrated in the technology- leading countries. Yet, a current CNFC−SFR does not assure meeting the INPRO acceptance
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limit regarding cost of energy. Capital costs for the first-of-a-kind fast reactors were two-three times higher than capital costs of thermal reactors. Since the nuclear cost structure is driven by up-front capital costs, the relatively high capital cost of fast reactors make it necessary to continue carrying out intensive R&D with an aim to reach acceptable values of capital costs. As demonstrated by French results, more than 5−15 % of fast reactor capital costs can be compensated by cheaper fuel costs compared to a thermal reactor (Fig. 6.2).
FIG. 6.2. Difference in cost breakdown between thermal (EPR) and European fast reactor (EFR). In spite of the expected positive effect of reduced fuel cycle cost of fast reactors on the total cost of electricity production, reducing the specific capital cost for fast reactors per unit of installed capacity is a task of first priority towards improving the economics of the INS. The survey of R&D being implemented by some countries of the Joint Study has clearly demonstrated that capital cost for fast reactors can be significantly cut. Among the most effective technical solutions identified were: reduction of steel consumption through designs with fewer loops and less required thickness for vessels and piping, elimination or size reducing of reactor systems, compacting plant layout, increasing thermal efficiency of the power conversion cycle, etc. (Fig. 6.3). Along with an examination of measures directed decreasing capital costs, a variety of R&D focused on the reduction of operational costs was considered. In case of India, the unit energy cost (UEC) is expected to be decreased almost two times (Fig. 6.4) due to identified R&D. The expected improvements of the INS economics via identified R&D were taken into account in development of the generalized reference model of the enhanced near term INS CNFC−FR. In turn, assessments of electricity cost produced by the INS CNFC−FR were made in the study for national systems with the closest fit to the generalized reference model. Results of the assessments using the levelized unit energy cost (LUEC) approach and evaluation of INPRO economic criterion on cost of energy by different countries of the Joint Study are compiled in Table 6.5.
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Scope for Capital Cost Reduction: Scope for Capital Cost Reduction: Japanese Approach Russian Approach • Less steel consumption: less no of loops, less thickness for • Extensive R&D leading to: vessels and piping Less steel consumption: less no of loops, less thickness for vessels and piping • Compact plant layout : less RCB and civil structures volume Compact plant layout : less RCB and civil structures volume Longer design life with higher capacity factor
Item Future JSFR BN-1800 BN GT-300 JSFR APWR BN-350 BN-600 BN-800 BN-600M BN-900 LWR (FOAK) Characteristic BN- BN- 5000 50 Electricity Output (MWe) 1500 1530 40 Power generation cost Mill/kWh ] 32.7 20.0 Number of Loops 2 4 800 1800 4000 30 pcfcmetal contentSpecific of reactor, t/MW(e) (design) 1,500 R/V (ton) 465 590 3 20 Net electrical output [MWe] 1,500 Building Volume, m /MWe 900 220 3000 13.7 IHX (ton) 576 (2 units) － 10.5 1760 (4 10 Fuel burn up [GWd/tHM] 60 115 2 8.2 9 SG (ton) 1000 (2 units) Construction Area, m /MWe 16 4 8 units) 2000 7 5.3 6 Plant efficiency [%] 35 43 Total Weight (ton) 2041 1900 2350 3 4.5 5 90 (R/V+IHX+SG) (APWRx0.87) Reinforced Concrete, m /MWe 200 50 4 Load factor [%] 95 1500 3 3 iccapital (design)$/kW cost, Thickness of R/V Wall (mm) 30 255 f
Plant life [years] 60 60 1217 2 Thickness of Piping Wall Speci 16 - 18 25 - 130 3 (mm) Steel Constructions, m /MWe 4.5 1 1000 1050 900 Discount rate [%] 22 920 Reactor Building Volume 890 1 800 0.9 3 150,000 360,000 0.8 Construction cost [$/kWe] 1,730 1,640 (m ) Metal Content of Reactor 9.7 4.4 700 0.7 700 0.6 Containment Facility Volume 600 0.5 Decommissioning cost [M$] 653 586 3 18,000 70,000 Facility, t/MWe 0.9 2345 (m ) 1
FIG. 6.3. R&D directed to fast reactor specific capital cost reduction.
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FIG. 6.4. Example of an approach reducing energy costs of a fast breeder reactor (Case of India). TABLE 6.5. RESULTS OF EVALUATION OF THE INPRO ECONOMIC CRITERION CR1.1 ON COST OF ELECTRICITY
CR1.1 Cost of Energy Joint Indicator Value Acceptance Study “k” IN1.1.1,CINS IN1.1.2,CA Limit Value Participant Value mills/kWh mills/kWh AL1.1.1 France 35.41 44.07 44.07 1 India 41.00 63.00−73.00 45.00 1 Japan 15.10 26.59 26.59 1 Rep. of 31.15 26.00−34.00 34.00 1(1.31) Korea Russian 17.74 24.50 24.50 1 Federation
As it follows from the results of nuclear energy cost evaluation (Table 6.5), expected values of electricity costs from all national near term INS CNFC−FR considered by the participants of the Joint Study are less by 8 to 47 % than the acceptance limits defined by the assessors. Thus, experience and knowledge accumulated give reason to judge that in the time span of 10−20 years cost improvements for a NPP with a sodium cooled fast reactor focused on
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design optimization and improved operational characteristics will result in achieving cost comparability with thermal reactor NPPs and fossil-fuelled power plants. The rather wide range of relative margins between the expected cost of nuclear energy in national INS and the corresponding INPRO acceptance limits (AL) can be explained by two main reasons. On the one hand, the cause of the discrepancy is a different approach for the selection of alternative energy sources. In case of France, the cost of electricity produced by a gas combined cycle (GCC) power plant was considered as AL (the value in Table 6.5 is converted from the original value 33.9 €/MWh of GCC turbines production cost by applying a coefficient of 1.3 as an exchange rate of Euro to US dollar). In India the cost of electricity produced by a 540 MW(e) PHWR unit and the costs expected from a 500 MW(e) prototype fast breeder reactor are considered as an acceptable level of electricity cost in the country. The cost of electricity generated by a typical steam turbine fossil-fuelled power plant was selected as AL in the Russian evaluation. The Republic of Korea used Generation-IV criteria and metrics for the values of alternative energy sources for comparison; the value of 26 mills/kWh corresponds to the upper bound of the ‘better than reference’ metric, the value of 34 mills/kWh reflects the upper bound of the ‘similar to reference’ metric (34.0–30.0 mills/kWh). In the Republic of Korea for the worst case the ratio “similar to reference to better than reference” produces a k value of 1.31. On the other hand, the wide range of discrepancy among the expected costs of nuclear energy in national INS and corresponding acceptance limits (AL) could be a consequence of differences in input data and the modes of economic calculations partly reflected in Table 6.4. Besides advanced sodium cooled CNFC-FR-MOX, other fast neutron systems based on heavy metal and gas coolants, novel fuels and technological innovations in CNFC may contribute to economic competitiveness provided their technical feasibility could be demonstrated. An envisaged increase of fossil fuel costs confirmed by today’s trends is in favour of long-term competitiveness for both thermal and fast reactor systems. Carbon emission credits could give additional incentives for nuclear power.
6.5.2. Ability to finance
INPRO user requirement UR2 is defined in [41] as: The total investment required to design, construct and commission innovative nuclear energy systems, including interest during construction, should be such that the necessary investment funds can be raised. Tables 6.6 and 6.7 demonstrate results of the assessment for user requirement UR2 consisting of criterion CR2.1 – financial figures of merit, and criterion CR2.2 – total investment.
TABLE 6.6. RESULTS OF EVALUATION OF INPRO ECONOMIC CRITERION CR2.1 ON FINANCIAL FIGURES OF MERIT
CR2.1 Financial Figures of Merit Joint Study Participant Indicator Value, IN2.1 Acceptance Limit Value, AL2.1
France IRR 14.0 % > 10 %/year ROI 14.0 % > 10 %/year India IRR 16.2 % ≥10.0 %/year ROI 25.5 % ≥10.0 %/year
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Japan IRR 19.0 % ≥10.0 %/year ROI 12.0 % ≥3.0 %/year Rep. of IRR 43.6 % ≥8.0 %/year Korea ROI 169.1 % ≥8.0 %/year Russian IRR 6.5 % ≥4.0 %/year Federation NPV 755 $M ≥ 0 In the INPRO methodology financial figures of merit are characterized by IRR – internal rate of return, ROI – return on investment, and NPV – net present value of cash flows.
TABLE 6.7. EVALUATION OF INPRO ECONOMIC CRITERION CR2.2 ON TOTAL INVESTMENT NORMALIZED TO A GW(e)
CR2.2 Total Investment Joint Study Indicator Value Acceptance Limit Value Judgment of INS Participant IN2.2 AL2.1 Potential $M $M France 2 486 * India 1330 < 1 490 YES Japan 1 615 < 2 596 YES Rep. of < 957 807 YES Korea Russian < 3 500 2 410 YES Federation * The value is converted from the original value of EPR total investment expressed in Euros (1 663 €/kW(e)) by applying a coefficient of 1.3 as an exchange rate of Euro to US dollar, and multiplying by 1.15 to reflect EFR total cost being estimated 15 % higher than that of EPR.
Results obtained by the participants of the Joint Study indicate compliance of INSs performance with the INPRO acceptance limits in this area of consideration. Being reduced to the specific value of capital cost, the latter varies in a relatively narrow range from $1225 per kW(e) in India to about $1660 per kW(e) in France. Of course, it should be kept in mind that the estimates depend on pricing policy and varying financial parameters, and should be treated as indicative values.
6.5.3. Risk of investment
INPRO user requirement UR3 states [41]: The risk of investment in innovative nuclear systems should be acceptable to investors taking into account the risk of investment in other energy projects. Table 6.8 compiles results of the assessment for the first criterion CR3.1 of UR3 – maturity of design. As it was already mentioned, the majority of the Joint Study countries have gained an extensive experience in operating experimental and demonstration fast reactors; most of the assessors also are among the recognized world leaders of nuclear power technology developers and user. Consequently, the results of the evaluation of criterion CR3.1 were predetermined (see Table 6.8). The second criterion CR3.2 of UR3 (construction schedule) requires providing evidence to the investor that the project constructions and commissioning times used in the financial analyses are realistic. The construction schedule should be assessed for one of three cases: deployment of first few NPP, deployment of a FOAK plant, and technology development.
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Since the majority of the countries participating in the Joint Study are interested in the development of the CNFC−FR technology, the judgment should be made on the potential of advanced techniques for reducing the construction time from concrete pouring to first criticality to the lowest practical value taking into account experience with previous NPP construction projects. Indicator values and acceptance limit values presented in Table 6.9 illustrate efforts of technology developers and designers aimed at achieving increased competitiveness of the INS CNFC−FR through reduction of construction times.
TABLE 6.8. RESULTS OF EVALUATION OF INPRO ECONOMIC CRITERION CR3.1 ON MATURITY OF DESIGN
Joint CR3.1 Maturity of Design Study Judgment Indicator Value, IN3.1 Acceptance Limit Value, AL3.1 Participant France Plan to address regulatory issues Design is licensable in country available and costs included in of origin. Plan to address development proposal regulatory issues available and YES costs included in development proposal India Design is licensable in country Design is licensable in country YES of origin of origin Japan Plan to address regulatory issues Plan to address regulatory available and costs included in issues available and costs YES development proposal included in development proposal Rep. of No licensing experience for SFR Plan to address regulatory Korea and related NFC facilities. issues available and costs NO Wealthy licensing experience included in development for PWR and CANDU (PHWR) proposal Russian Plants of same basic design have Plants of same basic design Federation been constructed and operated. have been constructed and YES Design is licensable in country operated. Design is licensable in of origin. country of origin.
TABLE 6.9. RESULTS OF EVALUATION OF INPRO ECONOMIC CRITERION CR3.2 ON CONSTRUCTION SCHEDULE
CR3.2 Construction Schedule (time of construction) Joint Study Indicator Value, IN3.2 Acceptance Limit Value, AL3.2 Participant years years France 4 ≤ 4.75 India 5 ≤ 4.5 Japan 7 ≤ 7 Rep. of ≤ 10 2.7 Korea Russian ≤ 7.5 6 Federation
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Table 6.9 shows that practically in all countries developing CNFC−FR technology projected time of the fast reactor construction meets national AL. The “financial robustness index of INS” − RI is the third criterion CR3.3 of UR3. The criterion CR3.3 requires a sensitivity analysis to demonstrate robustness of financial parameters to changes in postulated sets of circumstances. The indicator seeks to measure whether the INS has sufficient robustness to changes in assumptions, i.e. that INPRO acceptance limits will still be met under different (financial) market conditions. Results of the evaluation of criterion CR3.3 in accordance with the equations proposed in the INPRO economic manual [41] are compiled in Tables 6.10. As one can see, all national INSs meet the robustness criterion. At the same time, the indicator values in some cases are rather close to the acceptance limit. It is not a surprise since economics of the INS CNFC−FR is rather near to the critical range for several market indicators like investment required to deploy a given INS, period of investment payback, etc. Contrary to previous criteria of UR3, participants of the Joint Study did not have enough experience in using the robustness criterion to make a definite judgment based on the results of its assessment. Probably, future assessments will provide a more complete picture related to this aspect of the INS and will help to understand its robustness in comparison with alternative energy sources. TABLE 6.10. RESULTS OF EVALUATION OF INPRO ECONOMIC CRITERION CR3.3 ON ROBUSTNESS
Joint Study CR3.3 Robustness Participant Indicator Value, IN3.3 Acceptance Limit Value, AL3.3 France 1.5 India 1.53−1.78 Japan 1.64 Rep. of >1 1.00−1.34 Korea Russian 1.38 Federation
The last criterion of UR3 is “political climate” or political environment. The risk to loosing political and/or governmental support for further development and deployment of the INS CNFC−FR in the countries of the Joint Study that are realizing national programs in the area seems to be small (Table 6.11). Main grounds for continuation of the policy were discussed in the Chapter 3 of the report. From economic point of view, it is expected that investments that have been done in RD&D during past decades will begin to give return in the foreseeable future when the near term INS CNFC−FR will be commercialized.
TABLE 6.11. RESULTS OF EVALUATION OF INPRO ECONOMIC CRITERION CR3.4 ON POLITICAL ENVIRONMENT
Joint CR3.4 Political Environment Study Indicator Value, IN3.4 Acceptance Limit Value, AL3.4 Participant France Political support for long term nuclear Commitment power exists though political changes sufficient to enable a return remain possible on investment India Political support for long term nuclear
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power exists Japan There is mid- and long term policy of nuclear option Rep. of There is National long-term support for Korea development program Russian Political support for long term nuclear Federation power exists The economic part of INPRO methodology of basic principles, user requirements and criteria also includes a user requirement UR4 and corresponding criterion CR4.1 denoted as “flexibility”. With regard to this criterion it should be noted that in the present version of the INPRO manual [41] the criterion is not elaborated and no approaches to its evaluation are recommended. Due to these reasons, the Joint Study assessors did not carry out any study on this criterion. At the same time, qualitative conclusions made by the Joint Study participants allow to draw a conclusion that the INS CNFC−FR, which possesses such features like breeding capability and weak dependence of generation cost on uranium price, undoubtedly is flexible to accommodate to various market conditions and applications.
6.5.4. Conclusion on economic assessment
Technical feasibility of an INS based on a closed nuclear fuel cycle with a sodium fast reactor (CNFC−SFR) has been demonstrated in the technology-leading countries together with a vicinity of cost competitiveness. Yet, a current CNFC−SFR does not assure meeting the INPRO acceptance limit regarding cost of energy. Knowledge accumulated gives reason to conclude that within a time span of 10−20 years cost improvements for NPP with sodium cooled fast reactors focused on design simplification, increase of fuel burnup, small serial construction, etc., will result in an introduction of a near term advanced INS CNFC-FR-MOX with enhanced economic characteristics (Table 6.5) into the nuclear energy system of countries leading in fast reactor technology. Assessment of this near term advanced INS CNFC-FR-MOX performed by the countries of the Joint Study using the INPRO Methodology demonstrated that the INS, despite of differences in some characteristics stipulated by national approaches and priorities, meets national acceptance limits, thus allowing to claim overall compliance of such an INS with INPRO economic criteria, user requirements and basic principle. The INS is recognized by the countries of the Joint Study as a prospective system for sustainable energy supply with a very high potential for large scale deployment in the mid- to long term future. The overall conclusion of the economic assessment of the Joint Study is that the near term INS with a serial fast reactor based on available technologies will meet the INPRO economic basic principle: it will be affordable and available in 10−20 years in the countries mastering the fast reactor technology. These counties have gone a long way to reach the current status of this technology. The participants of the Joint Study conclude that countries with a sizeable nuclear power program have more incentive to raise the high total expenses and accept the risks in developing an INS CNFC−FR, than countries with a rather small nuclear power program. That’s why, most probably, the first phase of the system’s commercialization will be realized in the confined group of these countries with a sizable nuclear power program. Outcome of the assessment of INS CNFC−FR on economics is not restricted to only evaluation of the INPRO indicators and acceptance limits in the economic area of INPRO methodology. During the course of the assessment, participants of the Joint Study examined some particular economic issues related to reactor and fuel cycle technologies, and discussed
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ideas aimed at improving economic performance and competitiveness of the innovative system with a closed fuel cycle and fast reactors.
6.6. FEEDBACK ON THE METHODOLOGY IMPROVEMENT
The INPRO methodology for economic assessment in its present state is well formulated and documented. Based on the widely known approach of levelized discounted cost (levelized unit energy cost, LUEC), it allows to calculate all life cycle costs of the INS. One of the significant achievements of the approach proposed in the methodology is that it does not focus on contrasting economics of thermal and fast reactors but gives an opportunity to consider nuclear energy as an integral system with a high potential of competitiveness in the future energy mix. Two specific comments on methodology improvement in the economic area were provided by the participants. The first one refers to the criterion of total investment required to design, construct and commission INS (INPRO indicator IN2.2 of UR2). The detailed explanation given in the INPRO manual [41] on application of the criterion gives clear guidance to an assessor who is looking at deploying an INS rather than to an assessor who is determining whether to invest in developing of an INS or not. The principle difference between deploying an INS and its developing consist in the amount of investments to be put in RD&D. Even the leaders in mastering the fast reactor technology have not reached a state of its development where the use of a CNFC−FR is profitable enough to cover the cost of RD&D that is still needed to make a conclusive step to commercialization. Comprehensive considerations in the INPRO manual [19] of specific costs related to an innovative nuclear energy system in comparison with investments into a proven nuclear energy system would have helped to make a more substantiated judgment on the criterion by different groups of assessors. The second comment deals with models for economic calculations. The levelized unit energy cost (LUEC) model was proposed in the INPRO Manual for electricity cost calculations. However, the ability of this model to take into account specific features of nuclear power economics was disputed in Refs [42] and [43]. Participants of the Joint Study came to conclusion that the LUEC model does not reflect considerable economic advantages of nuclear power plants associated with their ability to operate over a long period of time (50−60 years) generating electricity of very low cost after 20−30 years of operation. This advantage of nuclear power − very important for consumers − is not covered by the LUEC model as a significant benefit since beyond two to three decades from the power plant start up the costs of money under a high discount rate become negligible, thus making the model insensitive to a long term vision. As far as just a long term horizon should be in the focus of a sustainability concept, it was judged by the Joint Study participants that the LUEC model is not an instrument absolutely consistent with the INPRO approach. Development of a more suitable model remains an important task for further efforts on the methodology improvement.
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CHAPTER 7 ENVIRONMENT AND RESOURCE SUSTAINABILITY
7.1. INTRODUCTION
The INPRO environmental basic principles for the assessment of INS [44] are focused on: • The acceptability of environmental effects, that means that the expected adverse environmental effects of the innovative nuclear energy system shall be well within the performance envelope of current nuclear energy systems delivering similar energy products, and • The fitness for purpose, that means that the INS shall be capable of contributing to the energy needs in the 21st century while making efficient use of non-renewable resources. It is generally admitted that the principal environmental burden of nuclear energy is linked to nuclear waste. The impact of production and different options to be handled are considered in the waste management chapter of the INPRO methodology and in Chapter 9 of the report. In this chapter we will focus on aspects quantified by stressors and corresponding indicators used for nuclear installations. Fast reactors, compared to present thermal reactors, have the advantage of having a higher thermal efficiency due to higher operating temperatures. Thus, most of the INPRO criteria which are normalized to the same energy output in terms of electricity production become favourable. Typically a thermal efficiency of 40 % for fast reactors compared to 33 % for present thermal reactors represents an improvement of more than 20 %. The experience gained from more than 40 years of operation of fast reactors in the world shows significant improvements compared to thermal reactors. Another important issue considered by the INPRO assessment is the use of natural resources. Fast reactors allow a much better use of uranium than present thermal reactors due to breeding of Pu in 238U. Some global figures illustrate this point: The IIASA B scenario considers 4500 GW(e) as the total nuclear installed capacity by 2100 (360 GW(e) in 2007). The use of thermal reactors to cope with this demand would consume 30 millions of tons of natural uranium. In case of using the present thermal reactors fleet with a partial renewal by other thermal reactors and finally a total renewal by fast reactors from 2040 on, the total natural uranium consumption will fall to 12 million tons, to be compared to the 15 million tons of proven conventional resources documented in the IAEA/OECD “Red book”. The radioactive content of waste generated by nuclear-power constitutes a potential risk for human beings and the environment. A key objective for ensuring nuclear energy sustainability is thus to implement safe and efficient waste management methods, particularly for long lived isotopes in high level radioactive waste (HLW).
HLW could be at least partially eliminated by way of partitioning and transmutation (P&T). This would facilitate design and optimisation of ultimate disposal of HLW, and its acceptance by the public, since the objective of P&T is to get rid of the more radiotoxic waste by incinerating it in fast reactors and ADS systems. However, HLW transmutation will not be easy to achieve, and could have a significant negative impact on nuclear energy cost, and since ultimate disposal is unavoidable and has, contrary to P&T, the potential to be a
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complete solution, it is very important to assess what could be the contribution, advantages and limitations of HLW transmutation, and to compare them to the direct disposal option. Among the criteria for selecting best options – such as safety considerations or balance of materials, energy and costs − environmental and health impacts should play a key role, because they are strongly connected to public acceptance and societal choices. We present here the environmental impact of a set of 5 nuclear scenarios selected to cover a very contrasted range of waste management options, in order to highlight and characterize the effect of HLW transmutation, and thus to estimate to what extent P&T could contribute to reduce the overall impact of long life waste on health and environment. We will also discuss the limitation of these results and of the methodology used, and suggest ways of improvement.
7.2. APPROACH USED
The approach is based on indicators obtained within three different and complementary steps: the nuclear scenario fuel cycle simulation, the scenario life cycle analysis (LCA), and the impact studies of each fuel cycle facilities involved in the scenario. The approach used provides assessment of the reviewed INS via a majority of the INPRO methodology indicators in the area of environment though it does not include evaluation for some indicators of the methodology.
7.2.1. Nuclear scenario neutronics calculation
This calculation is performed by using the CEA COSI fuel cycle simulation code. COSI allows simulating the operation of a nuclear fleet and the associated fuel cycle facilities over several decades. It is used for “nuclear power scenario” studies and calculates all fuel cycle material fluxes and inventories (mass balance, HLW production, waste radio-toxicity, glass canisters produced by TWh(e), etc.). Results obtained by COSI calculations allow to evaluate and analyse the consequences of choices made on the installed nuclear power and the nature of the reactor fleet, the types of the fuels, the characteristics of the various fuel cycle facilities (fuel fabrication and reprocessing plants, transport, interim storage, etc.), and the waste form. COSI models all types of reactors and fuels we needed for our studies: PWR of European pressurized reactor (EPR) type and fast reactors of European fast reactor (EFR) type with different breeding ratios. These are well known reactor technologies, in which transmutation has already been fully studied [45], with various types of UOX and MOX fuels, and also minor actinides (MA) transmutation fuels: americium targets for once through incineration in an EFR, or all MA mixed with plutonium (Pu) in EFR MOX fuel for complete trans uranium elements (TRU) recycling. Among the numerous information provided by COSI, we have selected very few significant quantities as indicators, all normalized to an energy production of one TWh(e): • Global mass of plutonium, americium and curium sent to the final disposal in kg/TWh(e). This quantity has been chosen because these actinides represent the bulk of waste radiotoxicity, and the Pu+Am+Cm mass reduction ratio with respect to an open fuel cycle is a very good indicator of the efficiency of transmutation; • The spent fuel assemblies’ number and volume (in m3/TWh) either for storage or final disposal;
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• The vitrified waste canisters number and volume (in m3/TWh) either for storage or final disposal; • The compacted waste canisters number and volume (in m3/TWh either for storage or final disposal. The last three indicators refer to the number and volume of spent fuel assemblies and waste canisters to fabricate, and to manage them in a long term interim storage facility for cooling, and prior to eventually sending them into a ultimate repository.
7.2.2. Life cycle analysis
Life Cycle Analysis (LCA) is the most recognized environmental assessment technique for a product (in our case, a normalized amount of electrical energy produced, i.e. 1 TWh(e)). LCA main steps consist of an exhaustive inventory of all flows of materials and energy needed for this product, followed by the quantification of potential impact indicators, related to major environmental stakes. LCA inventory needs a preliminary identification of all the steps involved in the production process (from mining to final repository in our case). For each of these steps, the input and output flows are identified: energy, natural resources, intermediate products and water consumption, emissions into water, air and soil, waste produced. All the data collected are then compiled in what we call the life cycle inventory. This data gathering has been performed using the EDF data base, issued to a large extent from the compilation of data of bibliographic origin (for instance, energy and raw material needed to fabricate a ton of concrete) and completed by specific data on French fuel cycle facilities, nuclear fuels and reactors. In the case of innovative and prospective facilities, fuel or reactors, like advanced reprocessing, americium transmutation targets, or fast breeder reactors, which do not exist at an industrial scale, information has been gathered from feasibility studies carried out by EDF or the CEA. The life cycle inventory also requires a few key parameters characterizing the nuclear scenario, which are produced by the scenario neutronics simulation with COSI: isotopic mass balance of fuel fabrication or reprocessing, fuel enrichment, and proportions of various types of reactors in the nuclear reactor fleet. LCA indicators are calculated by aggregating the inventory flows through the whole process, using the TEAM® software. 11 indicators have been selected for our methodological approach, chosen for their reliability in spite of data uncertainties. Four of them are related to non-renewable natural resources consumption: oil, gas, coal and uranium. Six are related to various types of emission: CO2, CH4, N2O, NOx, SOx, and particles. These ten indicators are expressed in tons/TWh(e). Those related to the greenhouse effect, are expressed in CO2 equivalent ton per TWh(e), thus normalizing and aggregating all gaseous emission contributions to CO2.
7.2.3. Impact studies
LCA methodology is well adapted to assess natural resources consumption, substance emission and a few potential effects like greenhouse effect increase, and it facilitates comparisons between several nuclear power scenario options thanks to the normalization to the same amount of electricity produced. However, LCA indicators are related to global effects on the environment, while radiological impact can be assessed only at the local scale of the emission, by impact studies on each fuel cycle facility, taking into account all local
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data. Contrary to LCA global indicators, radiological impacts cannot be normalized to one TWh(e) produced; they are expressed in milli Sievert per year (mSv/a), and have to comply with limitations imposed by regulation on health impact. Radiological indicators have been restrained to the dose impact on the most exposed population, because this impact is available for all current fuel cycle steps, and can be extrapolated for the prospective ones, and also because the highest attention is focused (by the nuclear industry) on health impact on the population living near the facilities. They have been gathered in 11 indicators according to the type of nuclear facility: mines, conversion, enrichment, UOX fuel fabrication, MOX fuel fabrication, plants, reprocessing, HLW disposal, low level waste disposal, very low level waste disposal, and eventually interim storage. All have been collected before being compiled from existing French fuel cycle facilities, except from the innovative ones: advanced reprocessing and new types of reactor (e.g., fast breeder reactors), for which they have been obtained by analogy with existing facilities, using for extrapolation the material fluxes calculated by COSI, and HLW ultimate disposal, for which the figure is the constraint imposed by the regulatory safety authorities.
7.2.4. Discussion of the indicators
The application of the approach used in the Joint Study results in 31 indicators, giving a nuclear fuel cycle environmental impact assessment as complete and reliable as possible. The following Fig. 7.1 sums up the multi-criteria approach implemented in the study:
Potential for scenarios Environmental
discrimination assessment
+ Fuel cycle inventory: Scenario neutronics simulation
Environmental potential impacts: Scenario Life Cycle Analysis
Environmental & health local impacts: Impact Studies results for each facility +
FIG. 7.1. Multi-criteria approach implemented in the Joint Study.
These indicators do not have the same meaning in terms of health and environmental impact. Indicators issued from the nuclear scenario neutronics simulation are very useful to differentiate the various options tested with respect to transmutation efficiency, and the reduction of the HLW radio toxicity, but can hardly be linked to a health impact: transuranium elements (TRU) represent the bulk of radiotoxicity in the final disposal, but they do not migrate, and thus have very little potential to contaminate the environment, even after hundreds of thousands of years (at least, with the assumption of a normal evolution of the final disposal, and thus without any human intrusion). Thus it is very difficult to give an environmental and health meaning to these indicators. On the contrary, impact studies bring precise information on nuclear facilities health impact, but it is impossible to link the radiological local impacts to a quantity like a
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TWh(e), which depends on the whole nuclear fuel cycle, and not on a particular facility, and thus impact studies are of no real use for comparing the nuclear scenario options. LCA has an intermediate position: it allows comparing various options thanks to its global cycle approach, but its indicators cannot be precisely linked to an environmental effect, and represent only a relative potential risk.
7.2.5. Scenarios description
Five nuclear scenarios have been selected:
1. PWR fleet, fed by UOX fuel, in an open cycle (spent UOX fuel sent to the repository);
2. PWR fleet, with spent UOX fuel reprocessing, vitrification of fission products (FP) and minor actinides (MA), and Pu "mono"-recycling (spent MOX fuel sent to interim storage);
3. PWR fleet, with spent UOX and MOX fuel reprocessing, Pu recycling and vitrification of FP and MA. Pu is recycled in MOX assemblies, with a uranium support enriched in 235U (a few per cent at equilibrium), and in EPR loaded with 100 % of MOX fuel. All spent fuel assemblies are reprocessed in this scenario. At equilibrium, the fleet is composed of 74 % of UOX loaded PWR, and of 26 % of MOX loaded PWR; 4. Mixed fleet with 45 % of PWR, 55 % of fast reactors (FR) recycling Pu and incinerating americium in transmutation targets. Neptunium and curium are vitrified with the FP. Am target design has been optimised to incinerate at least 90 % of Am in a once through strategy [46]. This scenario eliminates most of americium, while limiting curium production at a level low enough to keep a good radiotoxicity reduction performance. 5. FR fleet recycling all MA together with plutonium (fully closed cycle). Only FPs are vitrified. This scenario is representative of a Generation-IV system, complying with sustainability thanks to plutonium breeding, and minimisation of HLW resulting from Pu and MA incineration. In these scenarios, only TRU incineration is taken into account, since they represent the bulk of nuclear waste activity, thermal release and radio toxicity after a few centuries of cooling. They cover a large range of TRU transmutation options, from the first one with no transmutation at all, playing a role of a reference scenario for rating and comparing the transmutation performances, to the fifth one, recycling all the TRU, the others corresponding to intermediate strategies. This selection allows testing the sensitivity of the environmental and health impact with respect to the extent of TRU recycling. All these scenarios are supposed to be in a “steady state”, which is a quite academic hypothesis. In order to be more realistic and also to verify if the steady state hypothesis is still acceptable at least on the first order, a sixth “dynamic” scenario has been studied (Fig. 7.2.): it simulates the evolution during one century from the actual French PWR fleet towards a 100 % fast breeder reactor fleet with a constant total power (60 GW(e)):
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FIG. 7.2. “Dynamic” scenario of the evolution during one century from the actual French PWR fleet towards a 100 % FBR fleet.
The initial phase of this dynamic scenario is the current French situation, with UOX spent assemblies reprocessing and Pu mono-recycling. Then, between 2020 and 2035, 50 % of the fleet (2 GW(e)/a) is replaced by PWR of EPR type, and Pu recycling is stopped in 2025. In a third phase between 2035 and 2050, the second half of the fleet is replaced by fast breeder reactors of the "EFR" kind (European fast reactor, developed in the 80ies) with the 2 GW(e)/a rate. From 2050 till 2080 the fleet remains unchanged, with 50 % EPR and 50 % fast breeder reactors. Then from 2080 till 2095, the ageing EPR (which life time is supposed to be equal to 60 years) are replaced by EFR, which will compose the entire nuclear park from 2095 on. This sixth scenario is similar to EDF prospective view of its nuclear fleet evolution during the 21st century, and has been accurately calculated [47]. Its first phase is similar to the scenario 2 (Pu mono-recycling), and its final phase to scenario 5 (EFR fleet) but without MA transmutation.
7.3. RESULTS
Experience received at operating FOAK nuclear systems with fast reactors as well as results of neutronics simulations and Life Cycle Analysis for innovative CNFC−FR systems confirm that the specific physical, technical and technological features of a CNFC−FR provide a minimal adverse effect on the environment and humans during operation. Heat releases for a fast reactor are essentially lower as compared with a thermal nuclear reactor though they need further reduction to reach a level comparable with most advanced fossil fuelled energy sources. Gas-aerosol releases, 131I, long-lived and short-lived nuclide releases are below the threshold of recording instruments; inert radioactive gas releases are mainly determined by experimental studies in "hot" chambers. Consequently, radiological impact expressed in mSv/a is much below admissible levels (Table 7.1). The experience gained also testifies that avoiding mining/enrichment steps in a CNFC reduces greenhouses gases (GHG) emission (CO2, CH4, N2O) as well as emission of air polluting gases (NOX, SOX) and harmful particles (Table1 7.1).
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TABLE 7.1: SUMMARY OF RESULTS OBTAINED FOR THE 6 SCENARIOS
Scenario Unit 1 2 3 4 5 6 2015− 2035− > Cycle neutronics simulation /TWh(e) 2035 2095 2095 Pu+Am+Cm sent to the waste kg 27.9 18.7 6.66 0.93 0.15 0.55 6.2 4.8 Spent fuel assembly number 1.025 0.45 - 0.13 - - - - Spent fuel assembly m3 5 0.9 - 0.034 - - - - Vitrified waste canister number - 1.64 1.85 1.59 1.49 0.83 2.73 1.49 Vitrified waste to interim storage m3 - 0.29 0.32 0.28 0.26 0.15 0.48 0.26 Vitrified waste to final disposal m3 - 0.65 0.73 0.63 0.59 0.33 1.08 0.59 Compacted waste canister number - 1.825 2.05 3.17 5.25 0.92 5.0 5.25 Compacted waste to interim storage m3 - 0.33 0.38 0.58 0.96 0.17 0.92 0.96 Compacted waste to final disposal m3 - 0.72 0.81 1.26 2.08 0 .37 1.98 2.08
Life Cycle Analysis /TWh(e) Oil consumption ton 600 600 600 400 100 600 350 100 Coal consumption ton 1 000 1 000 900 700 400 1 000 650 400 Natural gas consumption ton 400 400 400 200 100 400 250 100 Natural uranium consumption ton 23.4 20.9 20 10.6 1.7 22.2 12.5 1.7 CO eq. Greenhouse effect potential impact 2 4600 4600 4400 3000 1400 4600 2900 1400 ton
CO2 emission ton 4400 4400 4200 2900 1400 4400 2800 1400
CH4 emission ton 6 6 6 4 2 6 3,5 2
N2O emission ton 0.1 0.1 0.1 0.1 0.02 0.1 0.06 0.02 NOx emission ton 15 15 14 10 4 15 10 4 SOx emission ton 24 25 24 16 7 24 15 7 Particles emission ton 41 36 35 19 2 38 22 2 Impact studies Mines and processing mSv/a < 1 (in the less favourable case of realistic scenario) Pierrelatte level ~0.002 Pierrelatte 0.002 Conversion mSv/a - - Malvési level ~0.07 Malvési ~0.07 Eurodif level Enrichment mSv/a Eurodif level ~0.0006 - - ~0.0006 Romans level UO Fabrication mSv/a Romans level ~0.0003 - - X ~0.0003 MOX Fabrication mSv/a - Melox level ~10-5 Nuclear power plant mSv/a ~0.001 La Hague level = 0,01 Reprocessing and advanced mSv/a - Advanced reprocessing : < 1 for future facilities (current reprocessing regulation) + COGEMA commitment < 0,03 Low Level Waste storage mSv/a 2 10-3 in operating phase ; 8 10-3 in surveillance phase Very Low Level Waste storage mSv/a 9 10-3 operating phase ; 9 10-5 in surveillance phase HLW deep geological disposal mSv/a Long term impact < 0.25 (Safety Fundamental Rule) Interim storage mSv/a Glass canister storage at La Hague < 10-3
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The impact mentioned for waste disposal has resulted from conservative calculations and thus maximized the amount of radioactivity which could be released. Concerning HLW disposal, the conservative scenarios and calculations recently performed lead to a maximal impact less than the level prescribed by the French fundamental safety rule III.2.f. The 6th “dynamic” scenario, has been separated in its three major periods: 2015−2035 (100 % PWR fleet), 2035−2095 (on average, 50 % PWR – 50 % FBR fleet), and after 2095 (100 % FBR fleet).
7.3.1. Main conclusions on acceptability of expected adverse environmental effects induced by the assessed INS (INPRO environmental basic principle BP1)
The main conclusions that can be drawn from results obtained for the five steady state scenarios are as follows:
• Pu multi-recycling in PWR (scenario 3) reduces by a factor 4 the amount of Pu + Am + Cm put into the final disposal if compared to an open cycle (scenario 1), and thus the radiotoxicity by about the same factor. This limited performance is due to the fact that a significant amount of plutonium is transmuted by successive neutron captures or decaying of 241Pu into americium, and then into curium. Mono recycling leads to a more limited result, with a 1.5 reduction factor. • It is interesting to note that all other indicators for scenario 2 and 3 are very close: plutonium multi-recycling compared to mono-recycling would not reduce significantly natural resources consumption (it requires practically the same amount of uranium per TWh(e), owing to 235U enrichment of uranium support for the MOX fuel), the major reduction of about 10 % with respect to an open cycle being already obtained with Pu mono-recycling; Pollutants emissions, greenhouse effect indicator, and radiological impact indicators are also almost the same. • Pu and MA recycling in fast breeder reactors (scenario 5) shows the lowest neutronics simulation and LCA indicators, except from the number of compacted waste canisters and volume, which appear here bigger owing to fast breeder reactor fuel assembly structures, which are more voluminous than PWR ones (at least for EFR technology which we chose for our study). The Pu+Am+Cm reduction factor with respect to an open cycle is about 200, meaning than the HLW radiotoxicity would decrease below the level of natural uranium in less than 1000 years. • The reduction of the number of vitrified waste canisters and volume results from the best electricity conversion efficiency of fast breeder reactors: the same amount of electricity requires less thermal energy, thus less fissions, leading to less FP, which are the elements driving the HLW vitrification volume. • Moreover, scenario 5 features a significantly reduced greenhouse effect and pollutants emission indicators. These indicators are mainly driven by the front end of the fuel cycle, and more particularly by the enrichment step with gaseous diffusion technology, which is ruled out for breeders. Note that the on-going replacement of gaseous diffusion by centrifugation technology, which requests far less energy, will strongly reduce the gap between scenarios 1 to 3 and scenario 5. • On the other hand, we have to point out that if the relative discrepancies between the scenarios seem important, the level reached by the greenhouse effect indicator remain very low whatever the nuclear power scenario considered, if compared to fossil fuels, like illustrated in the Fig. 7.3:
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g CO /kWh 2 1000
500
0
Nuclear CC Gas Coal
FIG. 7.3. CO2 equivalent emissions in g/KWhe for various sources of electricity production: nuclear, natural gas and coal.
• The reduction of greenhouse gas emissions resulting from scenario 5 with respect to scenario 1 is about 1.3 million tons CO2 equivalent for the entire French fleet during a year (400 TWh(e) produced), and would represent only 0.3 % of the total greenhouse gas emissions in France.
• Scenario 4, with a fleet composed of 45 % of PWR and 55 % of fast breeder reactors, occupies an intermediate position between scenario 3 (PWR fleet) and scenario 5 (fast breeder reactor fleet), and most of the indicators can be directly deduced from these proportions. The most interesting point to note is that the Pu+Am+Cm reduction factor with respect to an open fuel cycle is equal to 30, which is quite a good performance for the only americium transmutation in a once through strategy.
• An important conclusion that can be drawn from all these results is that the main discrepancies between the scenarios studied, chosen to represent a wide range of transmutation options, are mainly due to the type of reactors and enrichment technology used, and any transmutation option implemented has by itself a limited and marginal impact, except from the HLW radiotoxic inventory, whose reduction cannot be directly linked to an improvement of health impact.
• This conclusion is reinforced by the fact that radiological indicators are always very low for a normal evolution of the facilities, whatever the fuel cycle facility and the scenario considered, and generally far below the current regulatory limitation of 1 mSv/a for public exposure. Even if some scenarios allow to eliminate some steps of the fuel cycle (for instance enrichment in the case of fast breeder reactors), these eliminations have no significant consequences on the global radiological impact of the nuclear fuel cycle, which remains too low for being a discriminating criterion. This result could seem deceptive for the methodology, but more positively it underlines the outstanding level of minimization of health consequences already reached by the nuclear industry.
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• Results obtained for the dynamic scenario are quite consistent with those obtained for the steady state scenarios, even if it is clear that they can’t be simply extrapolated from the initial state (which corresponds to scenario 2), and the final one (which is close to scenario 5, without the minor actinides transmutation), mainly because the reprocessing of MOX spent fuel assemblies accumulated during the first phase of the dynamic scenario cannot correctly be taken into account by the extrapolation: to obtain accurate information on the HLW composition and quantities, it is necessary to model precisely the transition phase, like we did with the COSI code. However, the main conclusions that are drawn from the “steady state” scenarios are not called into question, but on the contrary confirmed by the dynamic scenario simulation.
7.3.2. Main conclusions on resource sustainability and fitness for purpose (INPRO environmental basic principle BP2)
Of course, natural uranium consumption is drastically reduced in the 5-th scenario, due to plutonium breeding in fast breeder reactors. The CNFC−FR would at the end let increase the nuclear energy resource base a hundred times in comparison with the present system of thermal reactors operating in an open fuel cycle mode. Based on a practically inexhaustible natural resource, the CNFC−FR can be de facto considered as an energy resource suitable for a large scale global exploitation. Only with the use of fast reactor systems nuclear power will have the potential to become a significant player in an energy mix for global sustainable development. We identify as an important task for further studies in this area a comprehensive evaluation of fresh water use in the CNFC−FR and a potential to reduce the use of other valuable natural resources.
7.4. CURRENT LIMITATIONS OF THE APPROACH USED
The approach used in the study for environmental assessment of the INS CNFC−FR is based on a model of the entire nuclear fuel cycle. Nevertheless it was not possible to be exhaustive, partly for lack of available data, partly to limit the complexity of the overall problem. Therefore a few steps were ignored; among them the decommissioning of buildings and sites is surely important. Anyway, the results obtained by the methodology for comparing transmutation options do not significantly depend on the implementation of decommissioning. The method used does not take into account depleted or reprocessed uranium management, which was not a major stake with respect to our objective. Concerning the radiological impact, dose to workers was not considered; this indicator is a basic data for the facility design, and integrates as such the regulation: workers radiological dose level will remain at the very most equal to the level currently accepted, whatever future facility considered, the problem being one of cost, out of the range of our study. Some local impacts on health and environment have not been treated, such as the chemical toxicity of emissions, and their impact on fauna and flora. Nevertheless, nuclear industry has achieved important progress in this area, and detailed evaluations are under progress. Results already available justify admitting that, like for radiological impacts on population, limitations imposed by the regulation are well respected. Indeed, what should be considered as the main limitation of the current approach is the fact that it makes the assumption of a normal behaviour of all facilities, and therefore excludes any accident. This comment relates not only to the approach used in the study but to the guidance
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of the INPRO methodology as well. A normal behaviour implies that all safety specifications are respected, with no significant health impact, rendering difficulty to discriminate between various nuclear scenarios. But to integrate a risk assessment in an INPRO evaluation would require a specific and heavy development, taking into account all fuel cycle steps and facilities, with no guarantee of success considering the heterogeneity of the facilities and of the time scales associated to various HLW options, from transmutation to geological disposal.
7.5. FEEDBACK ON THE METHODOLOGY
An important conclusion of the study is that it is not possible, using the methodology developed for assessing of the environmental and health impact of a nuclear power scenario, to really discriminate among scenarios implementing very different levels of HLW transmutation, that are representative of the whole range of available options. This result is coherent with the assumption of normal behaviour of the fuel cycle facilities, complying with safety specifications. Main discrepancies between these scenarios concern the uranium consumption, which is strongly limited in case of fast breeder reactor deployment, and greenhouse gas emissions, which is reduced for a fast breeder reactor fleet, due to the stopping of uranium enrichment by gaseous diffusion. Anyway, this impact remains very low if compared to fossil energy production. However, other indicators – at the boundary of the scope of this work- may be sensitive to different scenarios. HLW transmutation would allow reduction of the radio-toxicity inventory and the thermal power in the final disposal, and could have a non-negligible impact on quality and performance of waste. But these indicators, well representative of reprocessing-recycling and transmutation efficiency, cannot be associated by the methodology focused on a health and environmental impact. It means that it is necessary to find other criteria in order to discriminate between HLW management options. As it was mentioned above, the specific approach used in the study does not cover the full spectrum of environmental indicators developed in the INPRO methodology. Improvements of the approach used in the study such as taking into account decommissioning, or the radio- toxicological impact on flora and fauna, would complete the methodology, and thus it could be useful to add them, but they are unlikely to change the conclusion that has been drawn from the current study. There is a greater incentive to implement accidental risks assessment in the methodology, in order to compare the probability of accident and its radiological impact on workers or population according to the HLW management options. But this methodological development is undoubtedly a challenging one, considering the complexity of the problem. A lot of documents are referred to in the environment chapter of the present INPRO manual [44] and results of studies are shown as examples in many pages, however, it is difficult for assessors to understand the manual without reading these references. The manual should be prepared so as to enable the assessment of potential conformity of the INS with the environmental protection. The manual should distinguish the items to be assessed according to their maturity level (site selected or not, preliminary, conceptual or final design/ construction/ operation etc.) in the development of INS, because the items to be assessed may be limited to an early design stage of INS. The relationship to the “Environment Impact Assessment” laws and acts, which should have been established in almost all of the INPRO member countries based on the ”Basic Environmental Law” (1993) of the United Nations, should be mentioned in the manual
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because the assessment items must be almost common between them. Concerning INPRO UR1.1, some of the items (i.e., society, quality of life, biodiversity, psychology, property values, politics and infrastructure) are very difficult to evaluate and the meaning of the evaluation is not clear. If necessary, the meaning of these items should be mentioned in a new manual. The current name of BP2 should be revised into an appropriate one to understand the content from the name, for example, “resource sustainability.” “environment,” “safety” and “waste management” are related closely, however, the boundary conditions among safety, environment and waste management should be cleared in the manual. What is the meaning of “off-normal events” (mentioned in the present version of manual on page 9)? If the “off-normal events” equals to “abnormal events,” it should be excluded from the environmental assessment because they are already considered as “abnormal events” in the safety analysis.
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CHAPTER 8 SAFETY OF NUCLEAR INSTALLATIONS
8.1. INTRODUCTION
Safety is one of the main issues which influence public acceptance of fast reactors with closed fuel cycle facilities CNFC−FR. Hence it is important to identify safety issues and recommend an appropriate R&D program in a comprehensive manner. Reduction in occupational and public radiation exposure is a universal safety goal. Safety features need to be implemented ensuring probability of a major radioactive release to public domain due to any event to be less than 10−7 (corresponding to residual risk) [48]. Passive safety features have to be incorporated wherever feasible after thorough validation. Many of the generic safety issues related to thermal reactors, such as criticality, radiation hazards and fire are also of relevance in CNFC−FR. In order to appreciate the specific safety issues in CNFC−FR, a comparison of the parameters relevant to safety in a typical light water reactor (thermal) and a sodium cooled fast reactor is presented in Table 8.1. TABLE 8.1. COMPARISON OF SAFETY RELATED PARAMETERS OF THERMAL AND SODIUM COOLED FAST REACTOR FUEL CYCLE FACILITIES
Parameter Thermal reactor and Fast Reactor and Fuel cycle fuel cycle facility facility (CNFC−FR)
Core arrangement Near-optimal Not in its most reactive reactive geometry geometry Void coefficients Coolant void Coolant void coefficient is coefficient is positive or at-best zero negative (overall) Stored energy in coolant High (PWR) Insignificant Chemical energy potential Low High: sodium-water; sodium- air Inherent emergency heat removal Low High (particularly in pool type capacity reactors) Burnup, GWd/t ~ 40 100 (typical - can be enhanced) Pu content in irradiated fuel, % 0.9 23.5 Total radioactive fission product X Ratio of thermal efficiency of inventory /GW(e) water reactor / fast reactor, times X (about 30 % lower) Specific radioactive fission product 3000 6000 inventory, immediately after shutdown, Ci/kg Coolant activation Tritium and Radioactive isotopes of sodium corrosion products and corrosion products In addition to generic safety issues in thermal reactors and its fuel cycles, sodium fire, higher levels of radioactivity during fuel handling and fuel reprocessing and higher losses of Pu in waste streams, are some of the specific safety concerns in CNFC−FR.
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In this chapter, challenges in the design of CNFC−FR with respect to safety, innovations, breakthroughs and related RD&D addressed by various member states are compiled [49].
8.2. SAFETY FEATURES AND CHALLENGES
8.2.1. Safety features
Fast reactors have many inherent and engineered safety features. These have been already demonstrated in various reactors. A few of them are highlighted below: Reactor core: The reactor core has overall negative reactivity feedback coefficients in case of a rise of power or temperature. The neutron flux distributions are stable which is beneficial from an operation point of view. There are no neutron poison effects of fission products, as in thermal reactors. Coolant In almost all liquid metal cooled fast reactors, designed and operated so far, liquid sodium is the coolant. The high boiling point of sodium that implies that sodium remains in liquid state up to a temperature of 1175 K and its excellent heat transfer characteristics bring many advantages for sodium as a coolant. A high boiling point permits a high operating temperature of the reactor, still ensuring sufficiently high margin to avoid boiling of the coolant under all e design basis events. A coolant, remaining in liquid state during operating conditions without increase of pressure, facilitates to choose a low design pressure for sodium systems and components, thereby demanding lower wall thickness of vessels and pipes. The excellent heat transfer properties provide high natural heat removal capability, particularly in the pool type concept. Finally, sodium is non-corrosive at a high purity level. Unique engineered safety features Reactor safety, incorporating certain passive elements in absorber rod drive mechanisms as well as in decay heat exchange circuits, can be improved by taking advantage of high thermal expansion coefficients and high temperature operation. A sensitive leak detection system can be engineered to justify a leak before break concept with austenitic stainless steel (ASS) as structural material. A reliable cold trap design can be employed to maintain a high purity of sodium. Robust in-vessel core catcher concepts are possible in view of the high thermal potential of sodium. By adopting a pool type concept, the advantages of sodium coolant can be exploited, apart from the inherent high energy absorption potential of large dimensioned ASS steel structures. By incorporating a safety vessel surrounding the main vessel, a complete loss of coolant is prevented and a sodium leak from the vessel can be detected efficiently. Further, by adopting double walls for sodium filled pipes, leak in a main pipe can be easily detected and direct contact of sodium with air can be avoided. By use of an intermediate heat removal system, the conventional side of the plant becomes practically non-radioactive.
8.2.2. Challenges
Unlike water reactors, the fast reactor core is not in its most reactive configuration. The core has a high density of power, which can lead to a positive void effect. The physical-chemical characteristics of sodium cause sodium water reactions, and chemical toxicity problems. There are difficulties for in-service inspection (ISI) of structures under sodium due to the opaqueness of sodium. Repairing circuits and components in a post-accidental situation call for very challenging technologies. The fast reactor components, i.e. large size thin walled
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shell structures, are very sensitive to earthquake loadings. Even though high breeding is not an essential feature, realisation of the full potential of fast reactors is not possible without high breeding. However, the requirement for enhanced safety and improved economy needs advanced materials and technologies which call for extensive experimental, modelling and simulation activities. High burn-up targeted for future fast reactors introduces higher levels of radioactivity during fuel handling and fuel reprocessing. These challenges are addressed in innovative reactor concepts which are discussed later.
8.3. REACTOR SAFETY PHILOSOPHY
The design objective is to make every identified sequence which can lead to a complete core melt, highly hypothetical. All the situations which can lead to important mechanical energy releases should be practically eliminated. The confinement, however, should be designed to offer sufficient resistance to a hypothetical release of mechanical energy. To achieve this objective, the design adopts a defense in depth philosophy which enables the definition of simple scenarios without any cliff edge effects, passive safety features in the core taking advantages of the presence of a sodium void at the upper part of the core, Doppler effects, core expansion behaviour, and shutdown and decay heat removal systems without calling for an intrinsic safety active device and preventive surveillance and in-service inspection (ISI) and repair provisions. Passive safety features are introduced carefully after confirmation. Sufficient grace time is provided until human actions is admissible during the time period required for recovery of operation performance of failed safety systems during beyond design basis accidents due to loss of off-site and on-site electric power supply of the power unit with simultaneous failure of all absorber rods and violation of the heat removal from the reactor to a final heat sink.
8.4. ROBUST DESIGN FEATURES
A robust design starts with choosing high quality materials, adopting established design, construction and inspection standards, and guides and methodologies. General guidelines for achieving a robust core design are: reduced power density of the reactor core, a possible minimum of stresses and a reduced core height. The reactor design should incorporate higher design margins for strength and seismic resistance, an in-vessel primary circuit purification system, a concrete reactor vault with a sealed liner serving as a guard vessel, a simplified fuel reloading technology with a minimum number of operations, the elimination of isolation valves in the secondary circuit loops, a detailed analysis with validated computer codes, and testing in simulated environment (sodium and temperature). Special design provisions, such as diversity in shutdown and decay heat removal systems, are introduced to meet the safety limits with adequate margins for design basis events, in order to prevent the occurrence of beyond design basis accidents. Further novel design features such as a re-criticality free core, an effective core catcher, and a containment are introduced for the management of beyond design basis accidents. A critical examination and consideration of experiences, rationalization of the design approach by deliberate adoption of the ALARP principle, reinforced treatment of severe plant conditions; and continuous improvement of the defence in depth implementation are some of features that demonstrate robustness of a design.8.4.1. Innovations towards enhanced safety in the core
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Some of the innovations to enhance the safety of the core being considered by international expert groups are: a lowest possible ratio of sodium void effect/Doppler effect, a reduced sodium volume fraction, a higher fuel volume fraction, a reduction of the height/diameter ratio, a moderate core size, an annular core, use of a metallic core, use of advanced cladding materials (15/15Ti, ODS, etc.) to gain higher margins against cladding rupture during transients, state-of-art core monitoring instrumentations and improved core diagnostic systems (neutron and acoustic noise registration),.
8.4.2. Sodium fire and sodium water reactions
The main objective is to practically eliminate sodium fires and sodium-water reactions, which may degrade the safety functions in an unacceptable manner and lead to unacceptable consequences to the environment. In order to achieve this, apart from the existing design features, reduction of welds in piping by reducing number of sodium loops, shortening of pipes with a high chromium content, double wall piping, double wall steam generator tubes, intermediate heat exchangers integrated with a pump, and state-of-art inspection and repair systems are being considered. Passive systems for self-extinguishing of sodium fires when sodium is leaking from circuits include sodium discharge into a pan with a hydro-lock and sodium discharge into a tank blanketed by inert gas. Sophisticated numerical simulation of sodium fire and sodium water reactions based on extensive experiments and validation are some additional innovative features envisaged in future designs.
8.4.3. Reliable and diverse shutdown systems
A few innovations are introduced in shutdown systems: in case of a power excursion, an automatic negative reactivity insertion device by way of an enhanced expansion of the core and shutdown systems; use of curie point electromagnets or magnet switches to de-energise the magnet holding of the absorber rods; hydraulically suspended rods which would lower automatically into the core in case of any loss of flow due to pump seizure or primary pipe rupture; and devices to arrest the uncontrolled withdrawal of absorber rods and self-actuating devices to shut down the reactor directly under class 4 power supply failure.
8.4.4. Decay heat removal systems
Pumps with a high mass inertia to ensure continuation of heat removal while waiting for the establishment of natural convection (electromagnetic pump), an enhanced grace time for safety actions, passive decay heat removal features, like the use of shape memory alloys for dampers, and the implementation of redundant circuits and minimization of risks of defaults by common mode are a few innovations being thought of by the designers worldwide.
8.4.5. Core catcher
Subsequent to partial or whole core melt, the molten core (debris) would be settled at the bottom of the main vessel. This may cause the rupture of the vessel, if it is not protected. For this reason, a core catcher is placed with the objective of maintaining stable conditions within the vessel. The core debris settled at the core catcher is called corium. Hence, the core catcher performs a very important safety function. The robust design of a core catcher involves definition of scenarios for core melt relocation, settling behaviour of debris within sodium, post-accident heat removal of corium and design analysis to demonstrate integrity of the main vessel.
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8.5. PROBABILISTIC SAFETY ANALYSIS
Confidence in safety analysis methods, postulated accident scenarios and hazards is raised through probabilistic safety assessment (PSA). PSA is carried out on the basis of practice acquired during systems reliability analysis using internationally established methodologies (IAEA and GOSATOMNADZOR guidelines), a validated software complex, and a data base on equipment reliability generated in the course of reactor operations worldwide.
8.6. SAFETY OF FUEL CYCLE FACILITIES
Fuel cycle facilities are characterized by a variety of physical and chemical treatments applied to a wide range of radioactive materials in the form of liquids, gas and solids. Fuel is not localized as in the case of reactors. Operator interventions are more frequent. Hence criticality issues warrant more attention in fuel cycle facilities compared to reactors. Chemical processes use strong acids resulting in high levels of corrosion, which in turn may lead to radioactive leaks restricting operation and maintenance of the plant. Hence control of corrosion is one of the main challenges. Major steps in the nuclear fuel cycle consist of chemical processing of fissile materials. These may result in hazards to workers and members of the public from releases of chemically toxic and corrosive materials. Strong chemical acids are used to chemically dissolve the spent fuel during reactor fuel element reprocessing, removing the cladding material and enabling the separation of the plutonium and uranium from the residual fission products. The residual fission products, which comprise approximately 99 % of the total radioactivity and toxicity in spent fuel, pose a significant radiological hazard in what is typically a complex chemical mixture. During the solvent extraction processes, strong acids and organic solvents are used to remove the plutonium and uranium from the slurries. These processes can generate toxic chemical by-products that must be sampled, monitored and controlled. It is important to recognize that unplanned releases of these chemicals may adversely affect safety controls and may lead to radioactivity release too. Red oil formation leading to runaway reaction and explosion is known to be caused by violent tributyl phosphate (TBP) - nitric acid reaction and subsequent rapid pressurization (typical place of occurrence: High level waste evaporator or Pu evaporator). Incidents such as red oil explosion need to be addressed. The safety issues discussed above are common to all fuel cycle facilities. In fast reactor fuel cycle facilities, criticality and radioactivity are of enhanced concern, due to higher plutonium content in the fuel. Higher burnup of the fuel results in larger radioactive inventories. Large scale release of radioactivity to public domain is of less probability in fuel cycle facilities, other than reactors. However safety analysis has to be carried out for off-normal events. Safety analysis tools for fast reactor fuel cycle facilities are not as well developed and validated as in fast reactors. A consensus on safety analysis methodologies for abnormal transients during operation (design basis accidents as well as beyond design basis accidents), should be evolved after detailed deliberations among experts of countries with operational experience. Seismicity has to be considered in safety analysis. Probabilistic safety analysis for fast reactors fuel cycle facilities is still in infancy and requires considerable development efforts.
8.6.1. Fuel fabrication
Main safety issues in fuel fabrication facilities are related to:
• Fire, due to the presence of H2, UF6 ;
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• Criticality in storage facilities, due to flooding and contamination; and • Increased radiation levels due to accumulation of radioactive powders on surfaces in glove boxes. Since fast reactor fuel fabrication involves handling of plutonium, explosions occurring in glove boxes (in mixed oxide fuel fabrication), due to loss of cooling, are of concern. Fire and explosion may lead to a loss of integrity of the containment and consequent release of radioactivity. Some of the innovations recommended are the use of high integrity, leak-tight glove boxes with improved double skin ventilation, advanced monitoring systems for prompt detection of leaks and extensive automation/ remotisation in operation and maintenance.
8.6.2. Fuel reprocessing facilities
Criticality accidents, accidental release of radioactive materials, fire and explosion are the main safety issues in reprocessing facilities [30]. Criticality control is a dominant safety issue due to the presence of fissile material in various dispersible forms and the possibilities of vigorous physical and chemical reactions. Presence of water, which is undesirable from criticality point of view, is also a matter of concern. Use of high concentrations of nitric acid and oxidizing/reducing agents in the process introduce problems of corrosion in the process equipment and piping, which translate into leakage of radioactivity. Flammable, combustible and explosive materials such as tri-butyl phosphate-organic solvent mixtures are used. Hence fire is one of the concerns. Red-oil formation leading to runaway reaction and explosion in locations such as high level waste evaporator is well documented [50]. Fire in HEPA filters in off-gas release systems is also possible. The occupational areas in the plants have high radioactive contamination potential. As fast reactor fuels are irradiated to high levels of burnup, radioactivity content in the fuels is high and radiation levels in the reprocessing plants are also higher. Criticality concerns are significant due to high plutonium content. To ensure criticality safety, use of safe geometries, and safe fissile material concentrations, mass and volume is adopted. To have a compact geometry without sacrificing safety, annular and cylindrical pulsed columns of safe geometry with improved packing are employed. The sub-criticality margin, presently pegged at > 0.05 can be increased to > 0.1, by addition of neutron poisons/absorbers. All process cells are designed to eliminate criticality for submerged conditions and fully reflected conditions. Criticality monitors, both neutron as well as gamma based detection systems are provided in the facilities where fissile material inventories are high. This is very important in CNFC−FR, since gamma radiation levels are very high in the fuel. Multiple and diverse systems, such as (1) continuous monitoring systems to provide the levels of liquids in process vessels; (2) temperature and radiation level monitoring systems in glove boxes and operating areas and (3) pressure drop measurements across filters help in early detection of abnormal conditions and thus enhance safety. As corrosion is one of the major issues in reprocessing plants, controlling corrosion of process tanks, vessels and equipment by proper selection of materials is important [51]. Currently American Iron and Steel Institute (AISI) type 304L stainless steel (SS) is normally used for the fabrication of above components such that corrosion loss is reduced to less than 18 mpy (0.45 mm/year). Materials with much greater resistance to corrosion such as the nitric acid grade (NAG) SS are also being used to obtain longer life of the components, as NAG SS shows corrosion rates less than 10 mpy (0.25 mm/year). Fire hazards are reduced by
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prevention of build-up of radiolytic hydrogen in waste tanks by ensuring adequate cooling and enough free board volume in tanks and vessels. PuO2 containers have to be designed with adequate cooling to withstand maximum temperature and pressure rise during long term storage. Possible presence of sodium on the surface and inside the fuel pins has to be considered to prevent sparking and incidental explosions after contact with aqueous solutions or moist atmospheres. To reduce radioactive release to environment, multiple levels of filtration of off-gas is desirable. Use of evaporators operating under reduced pressures and lower boiling temperatures would enhance safety. Enhanced automation and remotisation of all the process steps should be employed to minimize exposure of personnel to radiation. Choice of appropriate materials for welds and periodic online in-service inspection, aid in maintaining low frequency of leaks (10-6) due to failure of pipelines. More reliable and ultra-sensitive detection systems such as Pu-in air monitors with automatic alarms are employed for continuous air monitoring, for prompt detection of abnormal events. These provisions lead to reduction in frequency of occurrence of abnormal events and their prompt detection, thus enhancing safety. The frequency of a major release of radioactivity into the containment/confinement of a fuel cycle facility can be reduced to 10-7 by better process control and monitoring systems. In case of any deviation from normal operating states, robust monitoring systems to initiate automatic action and automatic intervention systems have to be provided to ensure corrective actions. With these systems, it should be possible to minimize need for human intervention. Hence fairly large grace times, of the order of sixty minutes could be achieved. In future, it may be possible to envisage a scenario of no human intervention. However this aspect needs further analysis by international experts. Material balance methods shall be in place to ensure that fissile material accumulation can be detected as early as possible and is minimized. Online NUMAC using neutron collars can be provided to minimize loss of fissile material. Some of the other innovations for enhancing safety in aqueous reprocessing are: • Co-precipitation of U-Pu oxide to minimise the number of process steps and hence the exposure levels; • Recovery and burning of minor actinides- Am, Np and Cm to reduce radiotoxicity of the high level waste for a long period of time; • Partitioning of heat generating nuclides such as 137Cs and 90Sr; • Separation of long lived fission products, thereby reducing the radioactivity in wastes; • Salt-free chemical process routes wherever possible to reduce salt content in waste streams. Safety is also enhanced by co-location of front end (e.g., fuel fabrication, NPP) and back end (reprocessing and waste management) facilities. This would have benefits through minimal transport and avoiding multiple handling of radioactive materials in different plants of the fuel cycle facility.
8.6.3. Alternate processes
Pyrochemical processes are being explored in the context of CNFC−FR. These processes use molten salts and molten alloys which are resistant to radiation. Hence short cooled fuels with high burnup can be reprocessed using these processes. The process volumes for a given amount of fuel are low compared to aqueous processes and hence the plant will be compact which also enables co-location of the reprocessing plant with the reactor and refabrication
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plant. The waste produced by these processes is in the form of solid and the volume of waste is less compared to PUREX process and hence waste management problems are less. It is estimated that the volume of the TRU waste generated from a pyrochemical reprocessing plant will be around 12 times less than that from an aqueous reprocessing plant. In the absence of aqueous reagents, criticality related problems are also less, which could enable the processing of Plutonium rich fuels. Besides these, there are some specific advantages of using the molten salt electro-refining process for metallic fuels: (i) in this process, the fuel is maintained in the metallic state throughout the process resulting in the reduction of the process steps; (ii) minor actinides are co-deposited along with the fuel materials and get recycled back into the reactor to enable burning them and at the same time producing energy. However, the pyrochemical reprocessing methods have the following disadvantages: (i) these are high temperature processes requiring inert atmosphere; hence they involve sophisticated technology; (ii) Relatively less experience with industrial application; quantitative recovery of all the actinides up to 99% is yet to be demonstrated; (iii) low separation efficiency; (iv) these are batch processes which limits the scale of industrial operation; (v) technology for the new waste forms has to be developed; and (vi) methods for nuclear material accounting have to be developed.
8.6.4. Human factors
The importance of the human factor for safe and reliable operation of CNFC−FR is recognized by everyone and is an issue that should be dealt with systematically. There are two perspectives of the human factor: On the one side, the operating staff is seen as a valuable resource that is playing an important role in plant operation, testing, maintenance and inspection, and sometimes compensating deficiencies in automatic systems. On the other side, the human intervention has also to be seen as a factor of disturbance and of limited reliability, the consequences of which have to be taken into account in the design of all plant systems and functions, to ensure a sufficient level of safety and availability of the plant. Enhanced automation and online instrumentation are employed to reduce human intervention. Possible contributions of human interventions to accident hazards: (1) errors in plant operation, testing or maintenance, liable to contribute to the failure of systems or to their unavailability or giving rise to an initiating event; (2) interventions in incident or accident situations, liable to influence the sequence of events with, on the one side, the capability to handle unforeseen situations, but, on the other side, to perform actions only with a limited reliability [2]. As a common principle, it has to be ensured that: functions, assigned to the operating staff, constitute consistent tasks and correspond to the abilities and strengths of the operating staff (e.g., appropriate degree of automation, appropriate number of tasks, appropriate sharing among centralized and local operating actions); and that the man-machine interface (i.e. control room, screen-based and conventional control means, processing of information to be presented to the operators) optimally supports the tasks of the operators and minimizes human error. For innovative systems, a human factors engineering program plan (HFEPP) is an essential part of the design, which will take into account: • Feedback of experience including a formal methodology; • A probabilistic safety analysis taking human error into account;
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• Use of adequate (and quantitative) models considering the causes of human error, which may assist to find appropriate design measures to avoid the causes and thus minimize human errors; • Visualization of the status of equipment of the plant (components, systems, etc.), the dynamics of the processes, the performance of the automated processes and their relation with the state of the plant, resulting in a guide for operator actions.
8.7. ASSESSMENT OF SAFETY IN CNFC−FR BY INPRO METHODOLOGY
8.7.1. Assessment of reactor safety
In the INPRO assessment of CNFC−FR safety, fast reactor safety features have been discussed by France, Japan, the Republic of Korea and the Russian Federation. User requirements are specified in the INPRO methodology for the assessment. As per these requirements, the innovative nuclear system installations shall incorporate enhanced defence- in-depth as a part of their fundamental safety approach and ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installations. INS should detect and intercept deviations from normal operational states in order to prevent anticipated operational occurrences from escalating to accident conditions. The frequency of occurrence of accidents should be reduced, consistent with the overall safety objectives. If an accident occurs, engineered safety features should be able to restore an installation of an INS to a controlled state, and subsequently (where relevant) to a safe shutdown state, and ensure the confinement of radioactive material. Reliance on human intervention should be minimal, and should only be required after some grace period. The frequency of a major release of radioactivity into the containment / confinement of an INS due to internal events should be reduced. Should a release occur, the consequences should be mitigated. A major release of radioactivity from INS should be prevented for all practical purposes, so that INS installations would not need relocation or evacuation measures outside the plant site, apart from those generic emergency measures developed for any industrial facility used for similar purpose. It is expected that an assessment is performed to demonstrate that the different levels of defence-in-depth are met and are more independent from each other than for existing systems. User requirements specified in the INPRO methodology are to be applied in the assessment [52]. Safe operation of installations of an INS should be supported by an improved Human Machine Interface (HMI) resulting from systematic application of human factors requirements to the design, construction, and operation and decommissioning. For the safety analysis, both deterministic and probabilistic methods should be used, where feasible, to ensure that a thorough and sufficient safety assessment is made. As the technology matures, “best estimate (plus uncertainty analysis)” approaches are useful to determine the real hazard, especially for limiting severe accidents.
8.7.2. Assessment by participants
All the INPRO safety indicators are not completely analysed by all the participants. Different reactor systems were used in the assessment as outlined in the following. China has presented a preliminary assessment on the conceptual design of a CFR-1000. With a pool type concept, sodium as coolant, small reactivity margins, low power density core, a natural convection mode of decay heat removal, passive shutdown and decay heat removal
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systems, a part of user requirements (INPRO user requirement UR1.1 of basic principle BP1) are assessed only qualitatively. French presentation covers in detail the safety improvements of a 3600 MW(th) capacity EFR and an innovative sodium cooled fast reactor. Design parameters on improved core design are reported, such as a decrease of the sodium void effect, decrease of the void/Doppler ratio by a factor 2−4 compared to EFR, etc. Further, innovative design features for the shutdown and decay heat removal systems and core catcher are highlighted. A certain break-through in core design for oxide, nitride and carbide fuels to achieve higher margins for the sodium void coefficients are reflected. Additionally, French studies elaborated proliferation issues in detail. India has applied the INPRO methodology to a CNFC−FR. Assessment by Indians of a CNFC−FR opined that, with the INPRO basic principles along with the respective users requirements, it is possible to assess the safety of an INS CNFC−FR in a comprehensive manner. Certain INPRO indicators are found to be duplicated (e.g., frequency of occurrences, consequences of events, etc.). With the absence of a common benchmark nuclear energy system, it is not possible to quantify some parameters in the present assessment (e.g., design simplicity, robustness, depth of R&D, safety criteria); only a relative assessment is possible. Grading should be assigned for the parameters representing the user requirements. Japan has assessed the Japanese sodium cooled fast reactor JSFR (with 1500 and 750 MW(e) capacity) systematically to a large extent. INPRO indicator IN1.1.1 robustness of design is assessed by various parameters such as a high quality of operation, a capability to inspect, a lesser expected frequency of failures and disturbances, a grace period until human actions > 30 min, a high thermal inertia to cope with transients, detection and interception of deviations from normal conditions, reduction of accident frequency, restoration to a controlled state, minimal reliance on human intervention, reduced major release frequency of radionuclides in the confinement (due to internal events), prevention of major release of radioactivity, no evacuation, improved man-machine interface, etc. Further, assessments are reported on inherently safe characteristics and passive systems, risk from radiation exposures, RD&D results available for safety assessment. The Japanese assessment concludes that the design concept of the JSFR is simple and robust, and has enhanced passive safety features, employed deterministic and probabilistic approach in the safety design and evaluation.The Russian Federation has applied the INPRO methodology to BN type reactors. Some of the parameters analysed by Russians are robustness of design (simplicity, margins), grace time until human actions are required, inertia to cope with transients, probability of occurrence, number of barriers maintained, probability of large release of radioactive materials to the environment, confidence in the safety analysis methods, accident scenarios and postulated hazards. Total domestic investments for development of the sodium-cooled fast-neutron reactors are about 12 billion $ in the Russian Federation. Experiments are planned to study components of a passive emergency protection systems, an emergency cool down system, a catcher for retaining reactor core debris after core damage, etc. Safety analysis has demonstrated that under (severe) beyond design basis accidents, the frequency of release of radioactive products into the environment for the BN-800 reactor is below 10-7 per reactor year. Simultaneously, it was shown that the effective radiological exposure of the public is below limits requiring evacuation. It is observed that each country has considered its own model of fast reactor for the INPRO assessment, rather than the reference CNFC−FR. This is understandable, since every country has studied its fast reactor option thoroughly and is able to make definitive statements. However, inter-comparison of the assessments of the participants would be possible, if the reference configuration would be considered by all participants.
8.7.3. Assessment of safety of fuel cycle facilities
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Fast reactor fuel cycle facilities safety has not been discussed by any participating country other than India. This indicates that sharing of information on such fuel cycle facilities needs further emphasis. International collaboration and sharing of data, such as operational experience of these facilities with regard to safety are essential to harmonise the approaches to innovation and enhancement of safety of CNFC−FR in all countries. INPRO methodology for assessment [53] of CNFC−FR safety indicates that: (1) Assessment parameters for fast reactor fuel cycle facilities need to be specific to nuclear fuel cycle facilities and are not the same as for reactors. For example, in INPRO user requirement UR2.1, some of the defined sample indicators, e.g., available excess reactivity and reactivity feedback are not relevant for nuclear fuel cycle facilities. Also some parameters in UR1.5, e.g., release of radioactivity into confinement/containment, and in UR1.3, e.g., safe shutdown state, are not relevant for nuclear fuel cycle facilities. One needs to talk instead about prevention of re-criticality and contamination. (2) A lack of quantitative parameters for INPRO indicators and their acceptance limits is observed. Various initiating events must be identified and design basis events defined in a comprehensive manner. For these scenarios, acceptance limits for the indicators (for example, grace time) need to be developed by consensus. A consensus on safety analysis methodologies should emerge among various countries after detailed deliberations. Seismicity has to be considered in safety analysis. Probabilistic safety analysis for nuclear fuel cycle facilities needs to be developed and standardized. This requires considerable international efforts to render it as a viable tool for safety analysis.
8.7.4. Innovations for future sodium cooled fast reactors
Certain innovations are also considered for future innovative reactors. In core design, in order to eliminate the possibility of flow blockage and at the same time to achieve a lower pressure drop, power flattening is achieved without gagging by the use of different core zones with identical Pu content, with different pin sizes or adopting a design of sub-assemblies with perforated wrapper or without a wrapper by use of advanced spacer concepts. Some potential additional breakthroughs are reported: a stable power shape during burnup, a metal plate concept for metallic fuels in place of pellets, a ultra-long life core, a fuel assembly design for enhancement of molten fuel discharge after a unprotected core degradation (parallel path for molten fuel), and engineering for a stabilised sodium boiling in the upper part of the core region without voiding in the fissile part. In the natural decay heat removal circuit, multiple flow paths in the pool to facilitate increased natural circulation (e.g., via a thermal valve in the inner vessel) are proposed. As part of the shutdown system fusible shutdown devices are placed above the upper fissile zone, which should act when a fusible threshold is reached. The core catcher would have features to achieve enhanced cooling of debris and to prevent re-criticality; this can be achieved by incorporating an enlarged coolant plenum for molten fuel quenching and fragmenting the debris, a novel chimney for effective coolant circulations, a multi-layer tray for debris retention within a limited height for cooling and sub-critical state, well defined paths for debris, sacrificial layers, etc. Alternate coolants and a comprehensive in service inspection and repair strategy are also breakthroughs being thought of for a future design. 8.7.5. R&D in reactor safety proposed by various participants
French participants in the Joint Study proposed extensive R&D programme on core physics and technology design, preventive surveillance and in-service inspection (ISI) aspects related to safety.
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A few suggested R&D areas from the Indian side are as follows: a demonstration of the energy absorbing potential of structures hanging from the top shield, a computer application to plant maintenance, the development of sensors and repair of welds in sodium, plant dynamics studies to minimise the number of scrams and optimising the flywheel inertia, software verification and validation, a top shield design to minimise the sodium release to the reactor containment building (RCB), minimisation of fuel handling errors, introduction and validation of design features to make a core disruptive accident a residual risk, development and validation of sodium fire models in RCB and steam generator building (SGB), development of a cold trap to operate within a hot pool, numerical simulation of material behaviour under irradiation, fuel and clad mechanical interactions and core mechanics and development of PSA analysis methodologies at various levels a. Japanese participants in the Joint Study reported that RD&D is in progress for all innovative features introduced in the Japanese sodium cooled fast reactor (750 MW(e) and 1500 MW(e)), such as risk-informed approaches based on a proven data set, a reliability database CORDS for a fast reactor PSA, a sensitivity analysis of design parameters and the application of computer codes in particular to understand the scaling laws so as to decide upon the requirement of full scale testing.
8.7.6. R&D studies on safety of fuel cycle facilities
R&D on fuel fabrication looks to avoid wet processes and inflammable materials to enhance safety. Simplified pelletizing, sphere packing and vibro packing are some of the methods under consideration. R&D on novel process control, sensors for process monitoring and early warning systems is required to ensure a probability of major radioactive release to public domain due to any event to be < 10-7 (corresponding to a residual risk). On-line process control devices such as online alpha monitors can prevent letting in high concentration plutonium feed solutions into critically unsafe oxalate tanks and thus prevent criticality. The same device can also detect inadvertent release of alpha emitting plutonium/actinides into potentially active/ low active effluent streams. R&D related to improving corrosion resistance of equipment in reprocessing facilities is required. For components using higher concentrations of nitric acid, for example the dissolver of fast reactor reprocessing plants with 11.5 N nitric acid titanium is being studied, as it offers good corrosion resistance and corrosion loss is less than 5 mpy (0.13 mm/year). To overcome the corrosion of titanium in condensate zones of dissolvers, new alloys like Ti-5 % Ta-2 % Nb and Zircaloy-4 are being proposed, where corrosion loss can be reduced to 1 mpy (0.025 mm/year). To overcome the high corrosion behaviour of titanium with high iron content (> 0.075 % Fe), a novel DOCTOR (double oxide coating on titanium for reconditioning) coating process is being developed in India to improve the performance and service life. Corrosion monitoring techniques like electrochemical noise analysis are being explored for online corrosion monitoring of process vessels and waste storage tanks to provide reliable information on the progress of corrosion as it happens in the plant. Permanent monitoring of the integrity of the primary containment barrier by measuring contamination levels in the process cells ensures early detection of leaks. R&D on remote in-service inspection technologies is necessary to ensure minimum leaks and radioactive contamination. In-service inspection techniques like laser triangulation for thickness measurement of dissolver vessel, remote imaging for viewing leakage of waste storage tanks etc. are being developed for application in fast reactor reprocessing plants. Advanced non-destructive testing techniques like remote field eddy current, an integrated eddy current-giant magnetic resistance system,
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surface acoustic waves, etc., are being developed for unique application in assessing the quality of reprocessing plant components. R&D is required related to on-line/in-situ material balance methods for prompt detection of fissile material accumulation and online systems using devices such as neutron collars. Neutron collars are systems wherein a number of neutron detectors are placed in an annular cylinder (of moderating material such as high density polyethylene) surrounding the active pipelines carrying high level wastes. The neutron detectors are all coupled to increase the sensitivity of the system. In the case of fast reactor fuel cycle facilities, gamma activity is high due to a high burnup of fuel and techniques like active neutron interrogation are required for hull and waste monitoring. This is more so the case in fuel cycle facilities based on a thorium cycle, due to the presence of 232U, whose daughters are hard gamma emitters. In a pyrochemical process, development of simulated ceramic and metal based waste forms in order to characterise them for thermal and chemical stability, waste molecule intake, leaching behaviour, etc., is being attempted. Investigations are in progress to develop corrosion resistant materials and coatings for application in molten chloride environments in order to ensure the reliability of the equipment and vessels against corrosion failures. Development of human factor models, use of artificial intelligence for early diagnosis, real- time operator aids and automation for short-term accident management are required to enhance safety. Development of PSA methodologies for safety analysis of fuel cycle facilities, including identification of initiating events, probability assignments based on operating experience is one of the immediate tasks, which could be addressed by international efforts.
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CHAPTER 9 WASTE MANAGEMENT
9.1. INTRODUCTION
In the framework of this Joint Study on closed nuclear fuel cycle with fast reactors, the participants of the study have analysed how fast reactors can improve present conditions based on an open or semi-open fuel cycle with thermal reactors. The assessment was made on the basis of experience gained from the operation for several decades of sodium cooled fast reactors (BOR60, RAPSODIE, FBTR, PHENIX, BN350, BN600 and SUPER-PHENIX) and on the innovations considered in the national development programmes for industrial deployment during the next thirty years. The INPRO methodology basic principles and user requirements considered were: • Waste minimization (INPRO basic principle BP1 and user requirement UR1.1); • Waste management related to the protection of human health and environment, by meeting regulatory standards (BP2 and UR2.1 and UR2.2); • Waste management to avoid burden on future generations (BP3) by specifying end states for the waste and moving wastes to the end states as soon as practicable (UR3.1) and by including the cost of waste management activities in the estimated cost of energy from the INS (UR3.2); and • Waste classification to facilitate waste management and assure that all steps in the waste management process (BP4 and UR4.1 and UR4.2) are covered.
9.2. WASTE MINIMIZATION
Three key features of fast reactors allow a drastic minimization of waste production in comparison to other reactor types for the same amount of energy supply: • The use of liquid metal (sodium, lead) or gas coolants in fast reactors allows a higher temperature of operation leading to a higher thermal efficiency; • Fast reactors with a better neutron economy and physics characteristics offer a possibility to recycle long lived radioactive waste and to transmute it, and thereby reduce the total amount of waste; • The higher burn-up achievable in fast reactors will result in less production of high level waste per unit energy production. Transmutation can be done by fission or by neutron capture reactions [54]. In case of fission, the heavy nuclide is transformed into fission products, most of which are short lived (less than 50 years). In case of neutron capture, the heavy nuclide is transformed into another element that does not necessarily yield a significant reduction in the medium or long-term radiotoxicity. However, the elements produced by capture can be transmuted further by fission or capture. Thus, transmutation by successive captures towards higher elements with higher activities should be limited, but transmutation by fission is to be encouraged. Therefore, for the analysis of the physics of recycling, we evaluate two parameters: • Ratio α = σc / σf which indicates the probabilities of decay through capture rather than through fission at the first neutronic interaction; • Production of higher elements with curium (245Cm), which is representative of the successive neutronic captures and the efficiency of the transmutation.
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Table 9.1 presents the mean effective cross sections of capture and fission and their ratio, for the main isotopes of actinides and for two neutronic spectra: a typical PWR spectrum with a UO2 fuel, and a fast reactor spectrum.
TABLE 9.1. ACTINIDES FISSION AND CAPTURE NEUTRON CROSS SECTIONS AND σC / σF RATIO
Isotope Thermal neutrons spectrum Fast neutrons spectrum (PWR) (FR) σf σc α = σc / σf σf σc α = σc / σf 235U 38.8 8.7 0.22 1.98 0.57 0.29 238U 0.103 0.86 8.3 0.04 0.30 7.5 239Pu 102 58.7 0.6 1.86 0.56 0.3 237Np 0.52 33 63 0.32 1.7 5.3 241Am 1.1 110 100 0.27 2.0 7.4 243Am 0.44 49 111 0.21 1.8 8.6 242Cm 1.14 4.5 3.9 0.58 1.0 1.7 244Cm 1.0 16 16 0.42 0.6 1.4 245Cm 116 17 0.15 5.1 0.9 0.18
The capture/fission ratio is reduced by a factor of 5 to 10 when moving from a PWR spectrum (thermal or epithermal) to a fast neutrons system. Fast neutrons are therefore more efficient for transmuting minor actinides by direct fission. One example of a possible P/T strategy is represented in Fig. 9.1. Multiple recycling of minor actinides in fast reactors was studied in the French case [55] where the present reactor fleet is replaced in a first period by EPRs and by the second half of this century by fast reactors (Fig. 3.7 of Chapter 3). As shown in Table 9.2, the total amount of minor actinides can be limited to reasonable levels while the fission products are stored in glasses. Multiple recycling of minor actinides in fast reactors decreases drastically the potential radiotoxicity in the final repository as shown in Fig. 9.2.
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Partitioning -Reference Diagram Dissolver Solution
CO , H O, H PO Pu+U 2 2 3 4 Oxidative/Chem. PUREX Destruction Np Transmutation / recycle HLW An Np Transmutation Novel Solvents Spent TBP / recycle
An / Ln An & Ln Group Separation Bulk Separation Recovery of Cs & Sr Cs/Sr Am+Cm +Ln
Value Recovery Ln Value Recovery PGM (within DAE) PGMs & LLFP (Delay & Decay) Vitrification /Ceramic (Acid killing & Evap) High waste loading matrices (LLFPs)
Neptunium - 237
•Recovery of Np-237 (t1/2 : 2.1 million years) : To be addresses in the PUREX process. •Separated Np-237 to be stored under strict safegaurds (Critical Mass 55 Kgs) •Ground water release, transmutable preferably in fast reactors
FIG. 9.1. A possible partitioning and transmutation strategy with a CNFC−FR.
TABLE 9.2. MATERIAL INVENTORIES DURING THE CENTURY IN THE FRENCH CASE FOR 400 TWH(E)/YEAR
Material Dimension 2035 2050 2070 2100 Natural U t/a 7850 4200 4200 0 Natural U (Cumulated) kt 430 515 600 660 Enrichment M SWU/a 6 3.2 3.2 0 Pu t 455 576 685 848 Np t 22 28 30 22 Am t 50 61 65 47 Cm t 4 7 10 17 MA t 76 96 105 86 Trans-uranium (TRU) total t 531 672 790 934 Fuel with TRU in fleet % 0 50 50 100 TRU in storage t 27 28 29 30
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FIG. 9.2. Potential of reduction of radiotoxicity of the wastes disposed in a geological waste disposal (by removing Pu, MA and fission products). The rationale for the judgement on the INPRO criteria related to waste minimization is presented in Table 9.3. TABLE 9.3. EVALUATION OF INPRO BASIC PRINCIPLE BP1 WASTE MINIMIZATION
Basic Principle BP1:(Waste minimization: Generation of radioactive waste in an INS shall be kept in the minimum practicable
User Requirements Criteria INS value of Judgment of Potential Rationale for judgment Indicator (capability) Indicators Acceptance Limits UR1.1 (Reduction of waste at the Alpha-emitters & ALARP ALARP Very good Partitioning of MA will source): other long lived lead to significant The INS should be designed to radio-nuclides per reduction for long minimize the generation of GWa lived radionuclides waste at all stages, with Total activity per ALARP ALARP Very Good Separation of MA and emphasis on waste containing long-lived toxic GWa. long lived fission products can result components that would be in liquid waste with mobile in a repository environment. low activity Mass per GWa. ALARP ALARP Very good -do- Volume per GWa.@ ALARP ALARP Very good 1) Through adoption of reprocessing flow sheet that would avoid addition of extraneous reagents processes 2) Separation and concentration of MA and long -lived fission products. Chemically toxic ALARP ALARP Very good No extraneous chemical elements that would addition envisaged become part of the in reprocessing flow radioactive waste sheet per GWa.
Some key features to develop waste management strategies in CNFC−FR are summarized below in Table 9.4.
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TABLE 9.4. KEY FEATURES OF CNFC−FR WASTE MANAGEMENT STRATEGIES Waste Management for CNFC-FR
Waste Minimisation:
Partitioning of minor actinides
Separation of heat generating nuclides: Cs-137, Sr-90
Separation of selected long lived fission products (Tc, PGM)
Adoption of process flow sheet which minimises salt addition
Waste Immobilisation:
Development of Ceramic matrices with long term stability and higher capacity for waste loading
In conclusion, minimization of waste by recycling and reuse of fissile materials, which would otherwise be a high level alpha-emitting radioactive waste (HLW) containing long-lived toxic components, from the very dawn of nuclear power was a weighty argument in favour of the INS CNFC−FR development. By now, technical feasibility of MOX fuel recycling and vitrification of the HLW arisen from reprocessing of fast reactor fuel has been demonstrated at an industrial scale. In case of multi-recycling of fuel in fast reactors a 90 % mass reduction of fuel waste requiring repository disposal can be achieved that means a 5−7 times cut in the necessary number of repositories. The objective of waste minimization is an important driver for introduction of the INS CNFC−FR in all counties developing nuclear technology.
9.3. PROTECTION OF THE HUMAN HEALTH AND THE ENVIRONMENT
INPRO basic principle BP2 and its related user requirements, overlap with user requirements in the INPRO areas of safety and environment. Nevertheless, some trends can be given in this chapter based on relevant present experience. Inherent characteristics of fast reactors, essentially linked to the design, “blind” fuel and components handling result in a reduced exposure of humans to radiation and a decreased environmental impact. Industrial experience gained from the operation of PHENIX in France and other fast reactors in the world shows that the highest doses for workers are recorded during big components handling even if they are kept at extremely low levels. In 2006, the mean exposure in PHENIX was 0.05 mSv per worker and the maximum value 1.4 mSv. For comparison, the annual exposure to natural background radiation (not including medical exposures and human activities) is 2.5 mSv. The rationale for the judgement on the INPRO criteria related to the protection of human health and the environment is presented in Table 9.5.
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TABLE 9.5: EVALUATION OF INPRO BASIC PRINCIPLE BP2 PROTECTION OF HUMAN HEALTH AND THE ENVIRONMENT
Waste Management Basic Principle BP2: (Protection of human health and the environment) Radioactive waste in an INS shall be managed in such a way as to secure an acceptable level of protection for human health and the environment, regardless of the time or place at which impacts may occur. User Requirement Criteria INS value of Judgment of Rationale for Indicator Potential (capability) Judgment Indicator Acceptance Limits UR2.1: (Protection of Human Health) 2.1.1 Estimated dose rate to an 2.1.1 Meets Shall not Very good As a policy Exposure of humans to radiation and individual of the critical group regulatory exceed the exposure chemicals from INS waste management Standards current limits are on a systems should be below currently accepted 2.1.2 Radiological exposure 2.1.2 Meets regulatory reducing levels of workers regulatory limits trend. and protection of human health from Standards exposure to radiation and chemically toxic 2.1.3 Estimated concentrations 2.1.3 Meets substances should be optimised. of chemical toxins in working regulatory areas Standards
Waste management Basic Principle BP2: (Protection of human health and the environment) Radioactive waste in an INS shall be managed in such a way as to secure an acceptable level of protection for human health and the environment, regardless of the time or place at which impacts may occur. User Requirement Criteria INS value of Judgment of Rationale for Indicator Potential Judgment Indicator Acceptance (capability) Limits UR2.2: (Protection of the Environment) Estimated releases of Meets regulatory Shall not Very Good As a policy the The cumulative releases of radio-nuclides radionuclides and chemical toxins standards exceed exposure limits and chemical toxins from waste from waste management facilities current are on a management components of the INS should regulatory reducing trend. be optimized. limits
9.4. WASTE MANAGEMENT TO AVOID BURDEN ON FUTURE GENERATIONS
An inherent advantage of a CNFC-FR is less high level waste to be disposed per unit energy production compared to a thermal reactor system with an open fuel cycle. Yet, whatever the chosen fuel cycle, the long-lived components of waste will have to be put in a final waste form, packaged and the packages placed in some form of a repository. The integrated system of fast reactors with a closed fuel cycle will have to be demonstrated to be permanently safe according to current regulatory standards. The greatest emphasis today in national programs is to rely on underground repositories for radioactive waste. The designs and operation of these facilities vary, e.g., in the depth where packages are emplaced, the host geological medium chosen, and the period of monitoring prior to sealing and closure of the repository. Low -and intermediate-level waste packages are isolated in relatively near surface repositories in many states. The protective features include the waste packages, sealing materials in the repository, as well as the natural barriers to movement of radioactive material through the geological environment. Most advanced nuclear power countries are planning to dispose of spent fuel and/or high-level waste in deeper repositories in stable geological media. Although progress is being made, it has been difficult to site and license such a final repository, so no repository for this waste is yet in operation and instead interim storage facilities are used. Long-term safety of the final repositories could be improved by P&T involving the irradiation of long-lived radioisotopes to transform them into stable or short-lived elements. This could significantly reduce the total amount of long-lived radioactive material requiring final disposal. The rationale for the judgement of the INPRO criteria related burden on future generations is presented in Table 9.6.
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TABLE 9.6. EVALUATION OF INPRO BASIC PRINCIPLE BP3 BURDEN ON FUTURE GENERATIONS DUE TO RADIOACTIVE WASTE OF CNFC-FR
Waste Management Basic Principle BP3: (Burden on future generations) Radioactive waste in an INS shall be managed in such a way that it will not impose undue burdens on future generations. User Requirement Criteria INS value of Judgment of Rationale for Indicator Potential Judgment UR3.1 (End State): Indicator Acceptance Limits (capability) An achievable end state should be specified for each class of waste, which provides 3.1.1 Availability of 3.1.1 All required technology is Technology Very good Technology for permanent safety without any technology. currently available or reasonably expected to waste further modification. The expected to be available on a be available immobilisation Planned energy system should schedule compatible with the on schedule available for be such that the waste is schedule for introducing the waste from brought to this end state as proposed innovative fuel cycle. thermal reactor soon as reasonably fuel cycle; with practicable. The end state 3.1.2. Time required. 3.1.2 Any time required to bring Technology Very good few should be such that any the technology to the industrial expected to advancements release of hazardous materials scale is less than the time be available and to the environment will be specified to achieve the end on schedule modifications, it below that which is acceptable state. can be adopted today. for waste from 3.1.3 Availability of 3.1.3 Resources (funding, space, Resources Very good fast reactor fuel resources. capacity, etc.) available for available cycle. achieving the end state Partitioning of UR3.2 (Attribution of Waste compatible with the size and radionuclides Management Costs): growth rate of the energy system. and their value The costs of managing all recovery (in waste in the life cycle should 3.1.4 Safety of the end 3.1.4 Meets regulatory standards Meets Very good certain cases) be included in the estimated state (long-term expected regulatory will be in place cost of energy from the INS, in dose to an individual of the standards industrially. such a way as to cover the critical accumulated liability at any group). Waste is Very good stage of the life cycle. immobilized 3.1.5 Time to reach the 3.1.5 As short as reasonably as soon as end state. Specific line practicable. possible item in the cost estimate after its generation
As a conclusion it was stated that the INS CNFC−FR has a break-through potential to meet the fundamental principle of sustainability: not to impose undue burdens on future generations. Recycling of Pu and MA in the INS would decrease two-folds and more the level of radiotoxicity acceptable today for the disposal of waste. Eventually, radioactive waste from the INS fuel could be managed in such a way that its end state reaches radioactive equivalence with uranium ore, excavated for fresh fuel, thousands times earlier than in case of a OTFC. In this case the long term radioactive risk on future generations is practically excluded since integrity of radioactive barriers of the waste repositories can be guaranteed for the necessary time period.
9.5. WASTE OPTIMIZATION
The fourth INPRO basic principle BP4 refers to optimization of the waste management process with respect to overall operational and long-term safety. This will require a waste classification scheme that facilitates optimal management of various waste types within the INS. The rationale for the judgement on the INPRO criteria related burden on future generations is presented in Table 9.7.
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TABLE 9.7. EVALUATION OF INPRO BASIC PRINCIPLE BP4 WASTE OPTIMIZATION IN CNFC−FR
Waste Management Basic Principle BP4: (Waste optimization) Interactions and relationships among all waste generation and management steps shall be accounted for in the design of the INS, such that overall operational and long-term safety is optimized User Requirement Criteria INS value of Judgment of Rationale for Indicator Potential judgment Indicator Acceptance (capability) UR4.1 (Waste Classification): Classification scheme. The scheme permits Classification of the Very good Present The radioactive waste arising unambiguous, practical wastes exists. classifications from the INS should be segregation & These wastes can are adequate classified to facilitate waste measurement of waste be also classified for facilitating management in all parts of the generation. as per existing effective and INS. classification. safe waste management. UR4.2 (Pre-disposal Waste Time to produce the waste As short as reasonably Shall be followed Very good Capabilities Management): form specified for the end practicable. as per the safe exist. Intermediate steps between state. current practices. generation of the waste and the end state should be taken Technical indicators: Meets criteria as Meets criteria as Very good Adequate as early as reasonably Criticality compliance; Heat prescribed by prescribed by technology practicable. The design of the removal provisions; regulatory bodies. regulatory bodies. available to steps should ensure that all- Radioactive emission control address these important technical issues measures; Radiation concerns (e.g., heat removal, criticality protection; measures meeting the control, confinement of (shielding etc.); Volume / regulatory radioactive material) are activity reduction measures norms. addressed. The processes and waste forms. should not inhibit or Process descriptions that Complete chain of Capabilities to Very good Current complicate the achievement of encompass the entire waste processes from address all types of practices and the end state. life cycle. generation to final end wastes from the technologies state and sufficiently entire nuclear fuel with flexibility detailed to make cycle exist. for evident the feasibility improvisation of all steps and adoption of new technologies can safely address this issue.
Having a clear ultimate perspective in HLW management, the technology of CNFC−FR provides enough flexibility by an optimal use of resources (funding, space, capacity, etc.) for achieving a balance between long-term perspectives and short-term risks and costs. For a group of countries with a sizable fleet of LWR (France, Japan, Russian Federation) introduction of a CNFC−FR is considered as a way for avoiding the accumulation of plutonium in spent fuel of LWR and decreasing the amount of the spent fuel. Safe conditioning of ultimate waste (FP+MA) after reprocessing demonstrated in France will remain a practical option for decades until a new and better option for geological disposal based on partitioning and transmutation will become available. From the point of view of the participants of the Joint Study, a technological approach to waste optimization based on recycling and reuse of radioactive materials has a high potential not only for management of HLW from nuclear fuel but also for radioactive wastes arising from decommissioning of the nuclear installations both of OTFC and CNFC.
9.6. RESEARCH AND DEVELOPMENT
A high potential of a CNFC−FR to cope with the waste management area was clearly identified; it was at least partially demonstrated by the feedback experience from existing sodium cooled reactors in the world and in particular in the countries participating in the Joint Study. Nevertheless, the industrial deployment of technological options needs to be optimized by means of national and international RD&D programs. Some main challenges to be faced are summarized in Table 9.8.
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TABLE 9.8. WASTE MANAGEMENT RD&D TARGETS FOR CNFC−FR
Waste Management : RD&D Targets Reprocessing of spent fuel Reduction of secondary waste. (inc. partitioning) Improved separation of recyclable nuclides. Interim Storage Increased safety of interim storage. Methods Transmutation Reduced long-lived radioactive components in HLW. Demonstration of transmutation technology. Waste immobilization Improvement in waste forms (chemical durability, mechanical stability, etc.). Geological Disposal Demonstration of disposal technologies. Enhanced understanding of hydro-geo-chemical transport processes. Improve long-term monitoring technologies. Long term human Assessment of risks associated with waste management factors analysis systems that require long-term institutional controls. Design-based comparisons Incorporation of safety of waste management and fuel of waste arising from reprocessing in the fuel cycle evaluations. proposed advanced reactors and fuel cycles
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CHAPTER 10 PROLIFERATION RESISTANCE
10.1. INTRODUCTION
Public acceptance will be crucial in permitting a large-scale deployment of fast reactors with a closed fuel cycle. Public acceptance of nuclear fuel cycles is judged by their safety, radiotoxicity and proliferation resistance [56]. Consequently, the innovative nuclear energy system (INS) based on a closed nuclear fuel cycle with fast reactors (INS CNFC−FR) should include innovate features such as passive safety in the design of nuclear reactors, transmutation of long lived radioactive waste into short lived or stable nuclides and measures to prevent the proliferation of material or technology to be used to fabricate nuclear weapons of mass destruction. From the viewpoint of proliferation resistance, it is necessary to assess the INS and related fuel cycle technology in order to affirm proliferation resistance for their continued development. The proliferation resistance assessment should provide not only to the assessor the proliferation resistance of the INS under development but also to the developer some guidance on its enhancement. The proliferation resistance of a typical INS CNFC−FR defined jointly and outlined in Section 4.5 of the report is assessed by the participating countries, using the INPRO methodology [57]. Some feedbacks for improving the INPRO methodology and the user’s manual are also suggested in this chapter.
10.2. PROLIFERATION RESISTANT FUEL CYCLE
From the viewpoint of enhancing proliferation resistance, a future nuclear fuel cycle system and technologies should meet the following requirements [58]: • Attractiveness of nuclear material management: o increased fuel burnup; o recycling without separation of plutonium; o recycling with the extraction of fission products (dirty fuel or cleaner waste concept). • Accessibility, physical protection: o integrated cycle; o detection techniques and controls; o minimization of transports. • Safeguardability: o IAEA Safeguards, EURATOM, etc.
The first two requirements are related to intrinsic proliferation resistance features defined in the INPRO methodology for the proliferation resistance area and the third one is related to extrinsic proliferation resistance measures. Every effort to enhance proliferation resistance in future nuclear fuel cycle system and technology should identify technical solutions to meet these requirements. Potential technologies which can achieve an enhancement of proliferation resistance in fast reactors with a closed nuclear fuel cycle are featured by: • Decreasing the enrichment needs;
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• Avoiding the build-up of a plutonium mine; • Keeping fissile material within proper safeguards; • Non-isolation of plutonium. The technologies supporting the enhancement of proliferation resistance given above are already applied or being developed to implement specific intrinsic features in the design of fast reactors with a closed nuclear fuel cycle, which would also facilitate the implementation of extrinsic measures.
10.2.1. Fast reactors
Fast breeder reactors have no need of uranium enrichment. 238U is simply added into the main fuel to compensate for the burnup of this element in the reactor. Provision of the start-up fuel inventories for fast reactors allows clearing plutonium stockpiles and utilizing the spent fuel kept in ponds at nuclear power plants, especially as the radiation barrier to illicit uses of spent fuel declines with time. In this case, initial separation of plutonium and fabrication of fast reactor cores should take place at enterprises of nuclear states or at international nuclear technology centres. Transfer of the plutonium contained in spent fuel and kept in stockpiles to fast reactors and their fuel cycle facilities, enabling incomparably a better protection, will eventually block the channel of weapons material proliferation. All of the presently operating thermal reactors use proliferation resistant fuel, because the fuel consists of natural U or U enriched to below 20 % of 235U. The spent fuel of thermal reactors is also regarded as proliferation safe, regardless of its plutonium content of about 1 %. However, if the plutonium is separated out and recycled in thermal reactors as a very pure mixed oxide (MOX) fuel, then the fresh fuel becomes proliferation prone [56]. Most sodium cooled fast reactors use pure MOX fuel or a Zr based alloy of U and Pu. However, they also could use a fuel with enhanced proliferation resistant characteristics such as the one being developed for nuclear transmutation, i.e., Pu together with Np, Am, Cm and long lived fission products. Several fast reactor designs based on this technology are currently being considered. In the closed fuel cycle of the reference INS, 238U added to fuel during reprocessing is burnt up in the reactor, while plutonium, being an integral part of the fuel, circulates inside the highly active material. The present nuclear energy generation based on fission by thermal neutrons requires pure fuel to suppress parasitic neutron capture. But in a fast spectrum reactor, the presence of fission products is tolerated to a large extent. Thanks to the physics specific to fast reactors, fission products should be only moderately removed from fuel during reprocessing. Besides, Am, Np and some Cm can remain in the fuel and will subsequently be transmuted. Together, these impurities set up a high level of radioactivity and provide an intrinsic physical radiation protection of this fuel against theft. As a variant for reactor specifications, a fast reactor core with a breeding ratio (BR) of ~ 1.0 (also known as a breakeven reactor) [59] is being developed with emphasis on proliferation resistance. The fuel cycle of this reactor does not need plutonium separation. With a full plutonium production in the core, there is no need to use uranium blankets, which precludes production of excess plutonium and eliminates the need for plutonium extraction. For an equilibrium fuel composition, the fact that the spent fuel composition is very close to that of fresh fuel implies that plutonium is neither extracted nor added to the fuel. Fuel composition is adjusted by simply adding another portion of 238U into the main fuel to compensate for the burn-up of this component.
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10.2.2. Closed fuel cycles
With still abundant low price uranium ore, the once-through LWR fuel will remain the most economical choice for a nuclear energy system for a while. Whether the spent fuel of thermal reactors will be disposed directly or reprocessed depends on the availability of a reasonable alternative for electricity generation. The present spent fuel of thermal reactors is regarded as proliferation safe. Future processing of spent fuels will eventually be extended to the partitioning of minor actinides (Np, Am, and Cm) and certain long lived fission products together with plutonium. At present, the main obstacle to future large-scale deployment of nuclear power is the industrial method of irradiated fuel reprocessing relying on plutonium separation. The main feature required of a reprocessing technology meeting the principle of proliferation resistance should be that it leaves no room for plutonium separation from uranium everywhere in the process, which means that the two materials should always go together in a certain ratio and avoid an isolated form. Inseparability of uranium and plutonium should take its root in the processes and equipment used in reprocessing. For spent fuel reprocessing, there are aqueous reprocessing and pyroprocessing (or known as either pyrochemical process or dry reprocessing) methods under development. The traditional currently used aqueous reprocessing method is proliferation prone, because during the process a pure plutonium fraction is obtained and the process can be easily changed in this stage. Instead, for spent fuel reprocessing, some options in the reference INS CNFC−FR rely on an advanced aqueous process with minor actinides partitioning. In this advanced aqueous reprocessing, U and Pu oxides are co-processed at all times without uranium and plutonium separation. That is, no plutonium is isolated throughout the entire process. With this advanced reprocessing technology, spent fuel is processed and then re- fabricated in an integrated fuel cycle facility. The re-fabricated fuel is inserted into fast reactors. Minor actinides and fission products (e.g., Cs, Sr) can be recovered as valuables for selected purposes. To enhance proliferation resistance of aqueous reprocessing, an integrated recycling concept namely “global actinide recycle” [58] is being developed as a future nuclear fuel cycle system. In this form of recycling, fission products are the only and ultimate waste from natural uranium and depleted uranium, because actinides (uranium, plutonium and minor actinides) are all recycled. Additionally the addition of low grade plutonium into fresh blanket fuel as well as actinide co-recovery to limit access by high radiation levels in lower decontaminated and/or TRU fuel is being investigated. The advanced aqueous reprocessing brings advantages for enhancing nuclear non- proliferation, including a reduction of proliferation risk by minimization of transportation, recycling of non-isolated plutonium, decreasing enrichment needs and avoiding the build-up of a plutonium mine. As a variant for reprocessing, pyroprocessing (or dry reprocessing) is considered in the INS CNFC−FR. Plutonium is mixed at all times with uranium, other actinides, and fission products in the electrometallurgical treatment of U-Pu-Zr metal fuel. Therefore, nuclear materials in this closed fuel cycle cannot be used directly in weapons. Additionally, fission products can be blended into the final product, which gives a proliferation resistance close to that of spent fuel. The plutonium used in this cycle never reaches the chemical purity needed for weapons. The plutonium is extremely inaccessible, because at all times it is in an extremely radioactive environment behind thick shielding.
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Pyroprocessing systems are compact, and a fuel cycle facility can easily be collocated with the reactor, all but eliminating the need to transport nuclear fuel. Since the partitioning takes place in small batches with rather slow kinetics and because of its compactness and radiation resistance, the pyroprocessing is suited for an integral fast reactor concept based on a small- scale plant. The inherently heterogeneous batch processing can be advantageous because of unit accountability, but it could make precise measurement of the plutonium in the system difficult. However, because of the limited number of portals and movement of materials, containment and surveillance could be effective with proper safeguards. Compared with pyroprocessing, the advanced aqueous reprocessing is more advanced in its development and would render itself for a centrally operated large-scale plant. If such a plant is an international institution, it would meet proliferation concerns and at the same time guarantee nuclear fuel supply to states having no access to the reprocessing technology. It is noted that the means of incorporating safeguards technologies into the fuel cycle facility design should be developed in order to improve process transparency. As the fuel cycle technology develops, there should be a complementary activity to integrate technologies that will enhance extrinsic safeguards barriers.
10.3. ASSESSMENT OF PROLIFERATION RESISTANCE OF INS
In assessment of proliferation resistance of the INS CNFC−FR, that is, fast reactors with a closed fuel cycle, the effective closure of the fuel cycle with no U-Pu separation and the quantity of nuclear material in the fuel cycle are examined as main intrinsic proliferation resistance features which could provide benefits for proliferation resistance. Owing to various maturity levels of design and technology, very limited assessments are performed. Most indicators are evaluated in due consideration of the technical state and development vision towards future safeguards, proliferation resistance and physical protection for advanced fuel cycle. Within the Joint Study the Japanese program for developing a nuclear energy system of fast reactors with a closed fuel cycle (called FaCT) was assessed in detail using the INPRO methodology for this area. All INPRO criteria were found by the assessor to be fulfilled (see Annex B). In China, the conceptual data of the evaluated model were found acceptable, including the envisaged closed nuclear fuel cycle and INS of a CFR-1000 (China fast reactor as an INS model). A proliferation resistance (PR) assessment using the INPRO methodology for a closed nuclear fuel cycle and a CFR-1000 was carried out, and covered all principles and indicators of the methodology. This evaluation didn’t get a detailed conclusion because it is hard to determine any acceptance limit for evaluation parameters or for criteria in the INPRO methodology. Through the evaluation exercise in the PR area, the assessor mainly understood what key factors should be considered in the analysis of PR of a fuel cycle and a reactor. All assessments by the participants of the Joint Study confirmed that an adequate proliferation resistance can be achieved in the INS CNFC−FR, considering the progress and maturity of design development. The envisioned INS CNFC−FR relies on an active plutonium management based on a closed fuel cycle, in which fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The Joint Study concludes that due to realization of intrinsic features,
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the threat of proliferation in the closed nuclear fuel cycle with reprocessing does not exceed the threat of proliferation in an open nuclear fuel cycle with fuel enrichment. A comprehensive analysis is further required to examine intrinsic features that could contribute to enhancing the proliferation resistance of fast reactors with a closed fuel cycle. A key objective of the assessment of proliferation resistance of the INS would be to identify the elements of technology that still remain to be developed and demonstrated to enhance proliferation resistance. Technical responses may hamper misuse of a reactor or facility, but they are not able to prevent completely the misuse of a nuclear system, hinder the misuse of knowledge, or a break-out scenario. The role of technology development is to improve the effectiveness of a package of extrinsic measures and perhaps to ultimately provide a level of transparency that will help enable the expansion of nuclear energy by fulfilling sustainability requirements. But no technical arrangement, in absence of extrinsic measures, has been found in the study to block the possibility of misuse either of enrichment, or reprocessing facilities. Proliferation resistance can be achieved only by a combination of intrinsic features and extrinsic measures, and neither shall be considered sufficient by itself alone. Extrinsic measures such as a state’s commitments and obligations and the implementation of safeguards will remain essential. For States attempting a covert proliferation, facility characteristics and the ease of detecting diversion can be troublesome barriers. In this case, consideration of nuclear material (isotopic, chemical, radiological, detectability and mass and bulk) and technology (facility unattractiveness, accessibility, available fissile mass, diversion detectability, time required, and skills, expertise and knowledge) could be effective to check that an INS is (or not) an attractive mean to acquire fissile material for a nuclear weapons program. The host country is responsible for safeguarding and securing nuclear materials in their fuel cycle from sub-national or terrorist groups, through extrinsic measures such as access control and a protective force, and an effective nuclear materials accountancy program. Intrinsic features such as high radiation fields potentially can make this task easier and perhaps less costly. Analysis of an INS’ proliferation resistance should be performed from the early design stages on to operation. It is important to keep proliferation resistance in perspective with the INS development. Responsible governments must keep under control, to the extent possible, the know-how relevant to produce and process either highly enriched uranium (enrichment technologies) or plutonium (reprocessing technologies). New technologies of safeguards, monitoring, and inspections should be developed and applied. There are several parallels between the objectives of proliferation resistance and those of physical protection against unauthorized removal of nuclear materials. They are not discussed in this chapter. An example comprising judgments over relevant indicators is given in Annex A to the report. Political agreement in the area of non-proliferation settled by international treaties and commitments of States is the fundamental basis of all extrinsic measures both for open and closed fuel cycle systems. To support a non-proliferation regime, activities of the IAEA and other international institutes/agencies should be intensified. Countries with a developed nuclear power system and related infrastructure can provide assistance for newcomers, thus contributing to strengthening the global non-proliferation regime. Only the international community can ultimately improve the international political regime of non-proliferation and associated measures of control, protection and enforcement. Multinational fuel cycle centres (MFCC) may help coping with all listed problems to a large extend.
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The Joint Study emphasizes a key role of the INS CNFC−FR in creation of such MFCC. At the end-state of the closed cycle, material to be disposed contains a negligible concentration of fissile nuclei in an unattractive form for any kind of proliferator. This distinctive feature of CNFC in combination with the potential of fast reactors to economic reuse of spent fuel creates the necessary preconditions to take spent fuel of thermal reactors (TR) from countries with a TR fleet back to a MFCC after a short time in storage. This option promises to decrease for countries with small expectations in nuclear power deployment economic burden related to building of installations for intermediate storage and final disposal in an open cycle and simultaneously reduce the potential risk of proliferation during storage and disposal of spent fuel. In summary, the Joint Study makes the judgment that there is a very high potential of the INS CNFC−FR to meet requirements of the INPRO methodology in the area of proliferation resistance and, finally, to comply with the basic principle in this area to help ensure that INS will continue to be an unattractive means to acquire fissile material for a nuclear weapons program.
10.4. FEEDBACKS FOR IMPROVEMENT OF METHODOLOGY
The proliferation resistance of an INS CNFC−FR was evaluated by participating countries, using the INPRO methodology. This assessment partly also intended to improve the INPRO methodology for the area of proliferation resistance (PR). The intrinsic features and extrinsic measures are clear to understand and utilize for a judgement on the proliferation resistance of an INS CNFC−FR. However, the selection of reliable and commonly accepted values of the INPRO acceptance limits still needs further efforts. The evaluation of indicators dealing with multiple intrinsic features and extrinsic measures and their robustness, requires the establishment of comprehensive scenarios on acquisition paths for nuclear proliferation. It would be difficult for the assessor to determine them especially in case of an INS under development. In a cost effectiveness analysis of optimization in the design/engineering phase, it would be difficult especially for the designer to evaluate proliferation resistance quantitatively in a balanced manner for lack of detailed design information. From the assessment of PR of the INS CNFC−FR several findings and feedbacks for improving the methodology are additionally given as follows: • International societies should reach a kind of international standard (or consensus) of an effective acceptable level in proliferation resistance measures. To achieve this goal, it is suggested that the evaluation scales for intrinsic features should be changed, but not to be used as the acceptance limits; • Regarding INPRO indicators (IN2.1–IN2.4), numerical evaluation scales in the user’s manual should be replaced by more qualitative ones; • Regarding acceptance limits (AL3.4–AL3.6), it is recommended that the acceptance limit should be modified to “sufficient according to international practice”; • Regarding indicators (IN4.1 and IN4.2), the evaluation of multiple barriers and their robustness requires the establishment of comprehensive diversion scenarios for proliferation, which is difficult to determine in the case of an INS under development; • Regarding indicators (IN3.1–IN3.5, IN4.2), it is difficult for the assessor to evaluate proliferation resistance appropriately, especially, at very early stages of INS development including project planning because of the lack of detailed information;
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• Regarding acceptance limits (AL3.4–3.6), it is suggested the acceptance limit should be changed to “sufficient according to international practice”; • Regarding indicator IN5.2, it is recommended that “analysis is not yet done, but to be done” should be defined as “strong PR” at early stages of INS development.
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CHAPTER 11 INFRASTRUCTURE
11.1 INTRODUCTION
Issues other than technical requirements are important to potential users of an INS. Many of the factors that will either facilitate or obstruct the on–going deployment of nuclear power over the next fifty years are related to infrastructure – national, regional, and international. For a country which plans to deploy an INS, the INPRO infrastructure user’s manual [60] defines a basic principle, corresponding user requirements and criteria, and indicates potential developments and conditions that would facilitate the deployment of the INS. The infrastructure manual defines a methodology of infrastructure assessment. Like other INPRO manuals it presents indicators necessary for successful deployment and operation of an INS, and includes legal, institutional, industrial, economic and socio–political features (public acceptance). This chapter reports the results of the Joint Study of the INS CNFC−FR in the domain of infrastructure. The analysis of compliance of the INS characteristics with the user requirements in the area are presented in Sections 11.2−11.5 while detailed results of assessment via all INPRO indicators are displayed in Section 11.6 (Table 11.1).
11.2. DESCRIPTION OF THE CASE ASSESSED
INPRO infrastructure basic principle BP1 reads: “Regional and international arrangements shall provide options that enable any country that so wishes to adopt INS for the supply of energy and related products without making an excessive investment in national infrastructure”. BP1 consists of four user requirements (URs): UR1.1; legal and institutional infrastructure, (UR1.2); economical and industrial infrastructure, UR1.3; socio–political infrastructure (public acceptance), and UR1.4; infrastructure of human resources. Each user requirement sets up some indicators, evaluation parameters and an acceptance limit. For a sufficient legal and institutional infrastructure (UR1.1), a national legal framework should be established covering the issues of nuclear liability, safety and radiation protection, control of operation and security, and proliferation resistance. The legal framework should be adequately in accordance with international standards. Safety and radiation protection arrangements should also be established. For a sufficient economic and industrial infrastructure (UR1.2), the country planning to install an INS should adequately support the project during construction and operation. The size of the INS should match local needs, considering costs and benefits. For a sufficient socio–political infrastructure (public acceptance) (UR1.3), appropriate measures should be taken to achieve public acceptance of a planned INS installation. The public acceptance of nuclear power should foster the existence of a national energy policy, information and communication to the public as well as the participation of the public in the decision–making process. For an adequate infrastructure of human resources (UR1.4), the necessary human resources should be available to enable an operating organization to maintain a safety culture and achieve safe operation of the INS installation. The operating organization should have enough
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knowledge of the plant to be an intelligent customer and should keep a stable cadre of trained staff. The assessment of infrastructure is a unique area compared to other technical areas of the INPRO methodology. The assessment range covers requirements from the design stage through site selection, construction, to operation. This may produce a difficulty for the INS user to evaluate some indicators. However, the necessary requirements for a successful INS deployment should be able to be confirmed. The evaluation results of countries making first steps in building a nuclear infrastructure as well as of countries with a mature infrastructure should be instructive to obtain a wide range of evaluations. However, this Joint Study was performed only by countries with well- established nuclear power systems, and thus it was not able to obtain the above views. Nevertheless, evaluation results that benefited from experience with nuclear power generation are of interest and made it possible to present comments profitable for the methodology. Table 11.1 shows the status of nuclear power development in the Joint Study countries. TABLE 11.1 STATUS OF NUCLEAR POWER DEVELOPMENT IN PARTICIPATING COUNTRIES
Nation LWR or HWR FR Experience
Canada Operation China Operation France Operation RAPSODIE, PHENIX, SUPERPHENIX India Operation FBTR Japan Operation JOYO, MONJU Republic of Korea Operation Russian Federation Operation BR-1, BR-2,BR-5/10, BOR-60, BN-600 Ukraine Operation
In the Joint Study, the countries that presented evaluation results of their infrastructure were France, India, Republic of Korea, China and Japan. Most of the countries have experience with thermal reactor power generation and development of fast reactors. These countries also have a development plan of future INS CNFC−FR. A few countries which executed an infrastructure assessment have experience with electricity generation in fast reactors. Therefore, the nuclear system targeted for the assessment of infrastructure is the INS CNFC−FR that is generally outlined in the reference model (Section 4.5) and that is planned to be developed and deployed in the Joint Study countries.
11.3. DISCUSSION OF ASSESSMENT RESULTS
The assessment results of the five countries, i.e., France, India, Republic of Korea, China and Japan were summarized, and are shown in Table 11.2. The assessment results of each INPRO user requirement are discussed as follows.
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11.3.1. INPRO user requirement UR1 legal and institutional infrastructure
User requirement UR1.1 (legal and institutional infrastructure) is defined [60] as: Prior to deployment of an INS installation, the legal framework should be established to cover the issues of nuclear liability, safety and radiation protection, environmental protection, control of operation, waste management and decommissioning, security, and non-proliferation. The legal framework of nuclear law is adequately established and in accordance with international standards in all countries which presented results of the INS assessment. Safety and radiation protection arrangements are established. China The administrative regulation on safety supervision of civil nuclear facilities, the control regulation on nuclear materials and the administrative regulation on nuclear accident emergency of nuclear power plants in China were issued by the State council in 1986, 1987 and 1993, respectively. The national nuclear safety authority was established in 1983. Thereafter, having the relevant documents from the IAEA as reference, the national nuclear safety regulations, rules and guidelines have been compiled and issued. France Basic law of 1963 plus basic safety rules are established. European standards apply to the French nuclear system. India There is an independent regulatory system which oversees the safety aspects of nuclear reactors and fuel cycle facilities right from the design stage; it consists of the atomic energy regulatory board (AERB), a safety review committee for operating plants, and a project design safety committee. The acceptance limit of radiation exposures is less than stipulated by ICRP limits. Japan There exists an atomic energy basic law, a law on the regulation of nuclear source material, nuclear fuel material and reactors, a legal act on measures against nuclear hazards and safety guidelines by a nuclear safety commission. Republic of Korea The atomic energy act, radiation protection standards and a nuclear safety policy statement are established. Russian Federation The Russian Federation is party to the Vienna convention on civil liability for nuclear damage since 2005 and has a domestic nuclear insurance pool comprising 23 insurance companies covering liability of some $ 350 million. Industrial safety base in the Russian Federation is determined by a federal law, N 116, from July, 21st, 1997 and by regulations which define rules and require the receipt of Rostechnadzor’s approval for mechanisms and equipment, used on dangerous industrial objects.
11.3.2. INPRO user requirement UR2 economic and industrial infrastructure
User requirement UR2 (economic and industrial infrastructure) states [60]: The industrial and economic infrastructure of a country planning to install an INS installation should be adequate to support the project throughout the complete lifetime of the nuclear power
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program, including planning, construction, operation, decommissioning and related waste management activities. The availability of credit lines is sufficient to cover the INS project based on the results of each country’s assessment. Economic assessment indicates adequacy of adequate return of investment (ROI). Capacities of nuclear facilities are based on energy demand and supply prospects in the future. The companies which support the industrial infrastructure have experience of construction of nuclear power plants in the countries participating in the Joint Study. The assessment results of each country are explained as follows. China A rather complete national nuclear industry has been established in China, but it is still needed to introduce advanced nuclear engineering technologies from other countries and enhance the capability for meeting a larger supply of nuclear power with advanced technology in the near future. Based on experiences gained from the construction of past NPPs, a localization rate of 50 % for new 1000 MW(e) PWR and 70 % for 800 MW(e) PWR has been reached. Recently, domestic nuclear enterprises encouraged by the prospective trends of nuclear power installation development have started to enlarge their investment into manufacturing infrastructures and capacity. The localization rate could be raised in the near future. An adequate economic regime concerning a credit, insurance and tax policy has been established. France The results of the French study indicated acceptable values (see Annex A). India The availability of credit lines in India is based on investments for the project from financial institutions with debt /equity ratio of 30:70. In the ROI evaluation, at present, there is a deficit of 14 % in peak demand which is expected to increase. A large capacity size justifies the local energy needs and an increase in the per capita power consumption. The government of India encourages cooperation in the nuclear energy on mutually acceptable terms. All components for the nuclear industry can be manufactured within the country. Large industrial houses have the facilities and financial strength to take up manufacturing activities. India has roads, electrical grid and other infrastructures to meet the requirements of the nuclear industry. The national infrastructure growth is around 8 % which is sufficient for supporting nuclear installations. Japan The results of the cost-benefit analysis in the FaCT project showed that the investment for R&D brings in good returns through the commercialized fast reactor cycle deployment. The investment risk which was evaluated for the investment amount and pay-out time was found to be sufficiently manageable to cover the project. Japan has financing methods such as government subvention and federal aid for nuclear power. The road map of the future project for the JSFR authorized by government was worked out in the framework of the FaCT project. The FaCT project specified the nuclear power generation installed capacities based on energy demand and supply prospects and estimated the necessary size of installations. In the design analysis of the FaCT project, the technical feasibility and business applicability were confirmed with domestic manufacturers. The domestic manufacturers had adequate experience of construction of nuclear power plants.
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Republic of Korea The availability of necessary financing is not yet confirmed by the analysis of budget resources (government, industry and operator). The nuclear power program for an INS is currently incomplete. There is no fast reactor construction plan approved by the government and no fixed strategy for a closed fuel cycle. However, government is supposed to adequately support the project during construction and operation. The comprehensive nuclear energy promotion plan (CNEPP) defined a road map for a sodium cooled fast reactor development project that was authorized by government. The existing capabilities of domestic industry are well equipped for a domestic nuclear program. A cost-benefit study is performed by the government and industry in every planning phase. Russian Federation The results of the Russian study indicated acceptable values.
11.3.3 INPRO user requirement UR3 socio-political infrastructure (public acceptance)
User requirement UR3 (political support and public acceptance) is defined [60] as: Adequate measures should be taken to achieve public acceptance of a planned INS installation to enable a government policy commitment to support the deployment of INS to be made and then sustained. Each country participating in the Joint Study has a national energy policy, and is utilizing communication tools which provide knowledge about nuclear energy and information to the public such as a web page. In case of installation of nuclear facilities, for hearing the opinion of the residents and achieving acceptance, the public hearings should be performed. These action programs are useful to build a national consensus on the decision making process and foster public acceptance. However, there might be a risk that a serious accident or societal scandal will lose public acceptance. Some individual assessment results of each country are outlined as follows. China Encouraged by the safe and satisfying operation records of domestic NPPs, generally, in China there is public support to use nuclear energy for solving the difficulties of economic development due to lack of electricity. In the scheme of the national mid-long term science and technology development program (2006−2020) issued by the State council of the People’s Republic of China on February 9, 2006, the importance is recently realized to achieve mutual-understanding and support between a government’s decision and the public aspiration for nuclear energy utilization. The government issued orders via official documents of responsible organizations to force NPP owners to ensure safe operation of their facilities, continue the transparency of operation by informing the IAEA and transmitting nuclear knowledge to the public. The responsibility of the national nuclear safety authority designated by the government is to execute a serious supervision of NPP at each stage, and all important issues related to safety must be reported to the State council not only during operation but also in the construction or in the maintenance phase of a NPP. France Public acceptance is gained by organizations or institutions such as CSSIN (high council for nuclear safety public information), CLI (local information commission), CPDP (public debate commission), and the French national act on energy supply and atomic energy committee. The appropriate measures must be implemented to check public acceptance continuously. For
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example, in France, a periodic standardized poll is carried out under the responsibility of safety authorities. India Public support is available for the nuclear program due to realization that nuclear energy is essential for meeting the energy requirements of the citizens. Clear procedures exist for ascertaining public support through public hearings. Such hearings are mandatory before starting any new project. The public hearings are attended by representatives of local bodies, prominent citizens, and representatives of pollution control board and sectorial experts. The safety record of the Indian nuclear power program has been good with 250 reactor years of accident free operation. Some of the present nuclear installations have been running successfully for 38 years without any public interference. Currently, a very high number (7) of nuclear installations is under construction. Japan Some related policies such as an appropriate national energy strategy are established in Japan. For example, there is a new national energy strategy, a framework for nuclear energy policy, and an energy basic plan. A web page which provides support for education in the area of nuclear energy is set up to communicate benefits of nuclear power to the public. The documents of the administration are now open to the public based on the freedom of information act, e.g., the discussion of the governmental council is assumed to confirm the principle of openness to the public. Materials related to nuclear power are disclosed in general by the nuclear energy library, which tries to secure full transparency. Information about operation of nuclear facilities can be obtained from the web page of each electric power company. Common understanding of nuclear issues among the public administration, the nuclear industry and local residents, is developed by risk communication. In case of installing new nuclear facilities, listening to the opinion of residents and taking it into account is mandatory, and hearings should be held which are open to the public. Republic of Korea Adequate national energy policies, which are regularly revised, are available, for example, a basic plan of long term electricity supply and demand, a national long- and mid-term nuclear R&D program, and a comprehensive nuclear energy promotion plan (CNEPP). A communication system is well established to address public concerns. The information of nuclear facilities operation is presented in the web pages of electric power companies. The decision making process is structured in the form of a consensus by experts and ordinary people. Discussions are assumed to be a principle of opening to the public, and many public hearings related to installation of nuclear facilities are held. In case of policy formation, the government invites the public to offer their opinions and incorporates into the policy a wide range of views from the general public. A long-term commitment of the government covering nuclear waste disposal is clearly included in the legal framework. Technical details for nuclear waste disposal are under development. Adequate surveys are performed on a regular basis by nuclear industry. The results of domestic surveys show a positive trend by a majority of the public supporting the national nuclear program.
11.3.4. INPRO user requirement UR4 human resources
User requirement UR4 states [60]: The necessary human resources should be available to enable all responsible parties involved in a nuclear power program to achieve safe, secure
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and economical operation of the INS installations during their lifetime. The owners/operators should have enough knowledge of the INS to be intelligent customers and should keep a stable cadre of competent and trained staff. For build-up of human resources, learning opportunities about nuclear energy and training at public research organizations and universities in each country are provided. There are periodic safety and peer reviews by international organizations such as WANO for the assessment of the safety culture. The assessment results of each country participating in the Joint Study are as follows. China About ten universities have a faculty for nuclear engineering to complete the human resource cultivation meeting the needs of the national nuclear development program. A master degree and doctorate system for nuclear engineering has been established in some universities and nuclear institutes. The China institute of atomic energy has operated for years graduate schools to familiarize students with various nuclear subjects. The nuclear power plants also have established a local training centre with simulators or cooperate with other suitable centres to train plant operators and maintenance teams. In all types of activities, the emphasis is always put on safety culture. China has taken part in WANO and a peer review of NPP operation has been carried out. France In the area of safety culture, a peer review from IAEA is useful. Periodic inspections and audits by the safety authority are set up in France and a periodic (10 years) revision for a licensing renewal is performed in depth. Employment in the atomic industry contributes to benefits of society and security of supply is also maintained. India India has a large resource of educational institutions which train students in various disciplines. India’s educational system is one of the largest in the world, with 311 universities and around 16 000 colleges. The government offers career packages for attraction of talented youngsters to the task of developing nuclear technology. The department of atomic energy has run a training school for the last 50 years to tap the best talent from engineering and science. This has provided a sustained source of human resource for the department. There exist a number of collaborative schemes through which the expertise available in universities and educational and research institutions are exploited for R&D on nuclear energy. The atomic energy regulatory board (AERB) has a mandate for a periodic safety review process applicable to nuclear plants during its construction as well as operating phases. The plants are also open for a peer review by international organizations such as WANO. Japan The public research organization provides learning opportunities about nuclear energy and training. Human resource cultivation is advanced in cooperation with the local community and the university. For example, there is a nuclear technology and education centre (NuTEC), a nuclear power training centre, a BWR operator training centre, and a professional system of cooperation between graduate schools and licensed professional engineers. By nuclear related-industry’s personnel a nuclear safety network has been established. The main activity is the dissemination of information related to the safety culture of nuclear energy, a peer review among the members, and support of communication and education regarding nuclear safety. By efforts of the government and the atomic energy commission, a report on the ideal form of a nuclear safety culture was published, and an opinion exchange meeting was held on
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this report. Japan is participating in WANO. The main activities are providing accident or trouble reports, seminar hosting and peer reviews. Japan also participates in the activities of the forum for nuclear cooperation in Asia (FNCA), exchanging views and having communication about nuclear safety culture and human resources development. Republic of Korea The educational system in the Republic of Korea is well-established. For example, there is a nuclear human resource development centre, a nuclear power training centre, and a professional system of graduate schools and licensed professional engineers. Regional and international arrangements such as INTEC, ANET and WNU are concluded. At present, professional conditions and perspectives in nuclear industry are attractive compared with general industries. A management system for measurement, assessment and improvement is well established and shows favourable results. Participation in WANO activities can be a practice of exchanging views and opinions about safety culture.
11.4. FEEDBACK TO INPRO METHODOLOGY
11.4.1 Specific comments
Comments and feedback on the INPRO methodology [60] obtained from the Joint Study participants are described in the following.
• UR1.2 economic and industrial infrastructure Republic of Korea: There is a need for development of a quantitative method to determine benefits of a nuclear program.
• UR 1.3 socio-political infrastructure (public acceptance) Japan: It is suggested that the acceptance limit for the indicators IN1.3.1 and IN1.3.2 is “sufficient according to the international practice”, and not “sufficient according to the best international practice”. France: The acceptance limit on public acceptance seems difficult to apply.
• UR 1.4 infrastructure of human resources Japan: The acceptance limit AL1.4.3 requires cost estimation. However it is difficult to evaluate “Benefits to society” quantitatively. Only qualitative judgment is possible (IN1.4.3).
11.4.2 General comments
General comments are as follows: Japan, Republic of Korea: For a country which plans to construct an infrastructure from scratch, the methodology provides guidelines and a useful check list for constructing an infrastructure to assessors introducing nuclear power program for the first time.
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Japan: Most of the indicators evaluate the nuclear infrastructure condition in a country rather than the differences of INS. There is a difficulty to evaluate some indicators at the stage of project planning. (especially IN1.1.1, IN1.1.2, IN1.2.1, IN1.3.1, IN1.4.2, and IN1.4.3). Republic of Korea: There is a need for provision of enough examples and guidance achieved through international collaboration to assessors in developing or inexperienced countries for their successful assessment. For these points, the evaluation of several indicators will become more difficult for a country which has no experience with nuclear power generation. Therefore, a standard procedure should be provided and a revision of acceptance limits might be needed in the manual.
11.5. MAIN CHALLENGES OF INS IN THE AREA OF INFRASTRUCTURE
Infrastructure is a unique area in the INPRO methodology compared to other technical areas. Assessment of infrastructure is not an assessment of INS design concept but an assessment of infrastructure maturity in a country and globally. Thus, an infrastructure assessment has to be evaluated based on a broad view on activities by government, industry and public. Infrastructure development related to nuclear power generation significantly differs between an experienced and an inexperienced country. This section describes the main challenges which are extracted from the results of the Joint Study.
11.5.1. Infrastructure development
INPRO user requirement UR1.1 (legal and institutional infrastructure) and UR1.2 (economic and industrial infrastructure) are requirements on (the quality of) an infrastructure which should be developed in a country which plans to establish a nuclear infrastructure. Obviously, a country which has experience with nuclear power generation has already developed such an infrastructure. However, for a country which plans to build-up an infrastructure, establishment of a legal framework including a nuclear law is not facilitated by the INPRO methodology. For such a country, a global standardization of safety regulation and guidelines and of design criteria and design concepts would be more helpful for the effective development and deployment of INS. International companies should be established to provide nuclear fuels to the countries that need to operate the INS installation in the future, and more important international legal arrangements should be made to ensure a safe and reliable supply of fuels during the whole service life of the nuclear installations.
11.5.2. Public acceptance
INPRO user requirement UR1.3 (socio-political infrastructure) and UR1.4 (infrastructure of human resources) are important to enhance public acceptance. A country which has experience with nuclear power generation understands the importance of public acceptance. Government and nuclear primary sector of such a country have developed a system to enhance public acceptance and have spent a lot of effort to get public understanding. However, in the present situation, the public perception still needs to be improved regarding nuclear-related issues such as nuclear safety, public radiation exposure, treating/disposing of radioactive waste, plutonium handling, seismic safety, etc. It is important to enhance public
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acceptance rather than to develop merely a system of public acceptance. Therefore, the level of public acceptance should be checked continuously. Realization of innovative features with a high technical performance could also help to improve public acceptance.
11.5.3. Human resources
For INS CNFC−FR deployment, the human resource issue is very important. This issue could be different from country to country. Human resources in some countries are decreasing since for a long time in the area of nuclear power there was no sufficient inflow of young scientists and engineers that could provide a long-term perspective of the nuclear system development. On the other hand, in the case of a country which concentrates on nuclear energy development, a sufficient number of young scientists and engineers might be available and pursue a high quality of R&D. In any case, the cultivation and securement of human resources must be continued for nuclear development in the future.
11.5.4. International cooperation
Even if the infrastructure is insufficient for development or deployment of an NS in one country, it is probably available internationally. If the development of infrastructure is difficult for a country that plans to build-up an infrastructure, it might be realized by multilateral cooperation. International cooperation will help to enhance and complement a national infrastructure. A survey done in the Joint Study has showed complementarity of national conditions that creates promising prerequisites for mutually beneficial long term collaboration. No single country, taken separately, disposes of a full set of factors favouring development of CNFC−FR, i.e. high energy demand, high level of nuclear technology and infrastructure maturity, high resource of nuclear fuel, etc., while the global nuclear system does. Assurance of nuclear fuel supply and services is one of the most perspective areas for collaboration. An international human resource cultivation system will also be helpful for INS deployment. International legal and institutional infrastructure has to be significantly renovated to realise the potential of mutually beneficial collaboration. Participants of the Joint Study highly appreciate efforts to enhance a multilateral approach to the nuclear fuel cycle (MNFC) under strong IAEA safeguards. An INS CNFC−FR that is capable to convert spent fuel of thermal reactor into a valuable energy resource provides a principle opportunity to significantly extend fuel cycle front and back end services on a multilateral basis. Transition to a global nuclear infrastructure realizing the potential of synergy between fast and thermal reactors as well as the synergy between countries was acknowledged in the Joint Study as a qualitatively new stage of nuclear power development. This new stage will provide additional perspectives of growth for mature nuclear industries, facilitate the use of nuclear power for newcomers, and mitigate consequences of nuclear phase out in case of a radical change in energy policy. Thus, the INS CNFC−FR is judged by the Joint Study participants to be a critical component of nuclear power development in the future in accordance with the basic principle of INPRO methodology in the area of infrastructure: Regional and international arrangements shall provide options that enable any country that so wishes to adopt, maintain or enlarge an INS for the supply of energy and related products without making an excessive investment in national infrastructure.
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11.6. RESULTS OF THE INS EVALUATION IN THE AREA OF INFRASTRUCTURE
Table 11.2 presents results of the INS CNFC-FR evaluation in the INPRO area of infrastructure performed by different national teams of the Joint Study.
TABLE 11.2. COMPARISON OF ASSESSMENT RESULTS OF NATIONAL INFRASTRUCTURES
BP： Basic Principle BP1: Regional and international arrangements shall provide Options that enable any country that so wishes to adopt INS for the supply of energy and related products without making an excessive investment in national infrastructure. UR： User Requirement UR 1.1: Legal and institutional infrastructure
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN 1.1.1 EP1: completeness of In accordance [Acceptance] Legal nuclear law with international -All the countries of joint study have experience of framework EP2: adequacy of standards the nuclear power generation. established nuclear law -The legal framework of nuclear law is established and is adequate in accordance with international EP3: scope of standards. international legislation in the nuclear law EP4: completeness and adequacy of safety regulations and guidelines IN 1.1.2 EP1: independence of Safety and [Acceptance] Safety and judgment radiation -Safety and radiation protection arrangements are protection radiation EP2: functions of established. protection regulatory body arrangements established arrangements -Establishment of Activities of regulatory body established specially for INS is required.
UR： User Requirement UR 1.2: Economic and industrial infrastructure
CR： Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis P arameters Limit IN 1.2.1 EP1: credit lines for Sufficient to [Acceptance] Availability of infrastructure provided cover the project -Sufficient to cover the project from the results of credit lines by industry and each country’s assessment operator EP2: credit lines for infrastructure provided by government IN 1.2.2 (Not EPs identified) Adequate to [Acceptance] Demand for enable ROI -Economic assessm ent indicates adequacy to and price of ensure ROI energy products
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TABLE 11.2. COMPARISON OF ASSESSMENT RESULTS OF NATIONAL INFRASTRUCTURES (CONTINUED)
UR： User Requirement UR 1.2: Economic and industrial infrastructure (cont.)
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN 1.2.3 EP1: energy system Matches local [Acceptance] Size of expansion plan needs -Capacities of nuclear facilities are based on the installation EP2: size of nuclear energy demand and supply prospect in the future. fuel cycle facilities (other than NPP) IN 1.2.4 EP1: survey of the Internally or [Acceptance] Support existing capabilities of externally -The manufacturer has experience of construction infrastructure the national industry available of the nuclear power plant in some countries. EP2: plan for national -Internationally available. participation IN 1.2.5 EP1: cost benefit VNI>NII [Acceptance] Value of study performed by (NII: national
UR： User Requirement UR 1.3: Socio-political infrastructure
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.3.1 EP1: existence of a Sufficient [Acceptance] Information to national energy policy according to best -national energy policy. public international practice
-The communication tools, which provides knowledge about the nuclear energy and information to the public. -the web page in the government and electric companies.
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TABLE 11.2. COMPARISON OF ASSESSMENT RESULTS OF NATIONAL INFRASTRUCTURES (CONTINUED)
UR： User Requirement UR 1.3: Socio-political infrastructure
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.3.1 EP3: information on Sufficient [Acceptance] Information to the operation of according to best -Sufficient public nuclear facilities international practice
-Some concealments in the past.
UR： User Requirement UR 1.3: Socio-political infrastructure
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.3.1 EP4: addressing of Sufficient [Acceptance] Information to concerns raised by according to best -Sufficient public the public regarding international nuclear installations practice
-The discussion is assumed to be a principle of opening to the public
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TABLE 11.2. COMPARISON OF ASSESSMENT RESULTS OF NATIONAL INFRASTRUCTURES (CONTINUED)
UR： User Requirement UR 1.3: Socio-political infrastructure (cont.)
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.3.2 EP1: appropriate Sufficient [Acceptance] Participation of decision making according to best -Public hearing and comment the public in process international
UR： User Requirement UR 1.3: Socio-political infrastructure (cont.)
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.3.3 EP1: confirmation of Sufficient to [Acceptance] Long term long-term commitment enable ROI -Sufficient commitment by government
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TABLE 11.2. COMPARISON OF ASSESSMENT RESULTS OF NATIONAL INFRASTRUCTURES (CONTINUED)
UR： User Requirement UR 1.3: Socio-political infrastructure (cont.)
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.3.4 EP1: surveys about Sufficient to [Acceptance] Public public acceptance ensure there is -Public comment acceptance of performed on regular negligible -Hearing open to the public basis political or policy nuclear power -Risk communication risk to the investment -However, there will might be a risk.
EP2: adequacy of -Serious accident or societal scandal will lose survey public acceptance.
UR： User Requirement UR 1.4: Infrastructure of human resources
CR：Criteria IN： Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.4.1 EP1: educational and Sufficient [Acceptance] Availability of training system for according to -Established education system human manpower needed in international NP projects experience resources
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TABLE 11.2. COMPARISON OF ASSESSMENT RESULTS OF NATIONAL INFRASTRUCTURES (CONTINUED)
UR： U ser Requirem ent UR 1.4: Infrastructure of human resources (cont.)
CR： Criteria IN ： In dicator EP : Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.4.2 Assessm ent m ethod Presence of [A cceptance] E viden ce th at to determ ine the p eriod ic safe ty -Periodic Safety Review safety culture safety culture. review -W AN O P eer R eview mechanism, prevails -Periodic Safety Inspection [R ef.] covering all technical ・ A S C O T gu id e lin es infrastructure and ・ S afety C ulture m anagement Im provem ent Matrix areas (S C IM )
UR：User Requirement UR 1.4cInfrastructure of human resources (cont.)
CR：Criteria IN: Indicator EP: Evaluation AL： Acceptance Evaluation results and basis Parameters Limit IN1.4.3 EP1: skills BTS > cost [Acceptance] Benefits to improvement of necessary to
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CHAPTER 12 R&D AND COLLABORATIONS FOR INNOVATIVE CNFC-FR
12.1. INTRODUCTION
Fast reactors with a closed nuclear fuel cycle have to be economically attractive, safe, environmentally benign and proliferation resistant, in order to be a viable alternative to the conventional sources of power/energy. The Joint Study concluded that to achieve these objectives, a comprehensive programme of R&D on a variety of areas, with an inter-disciplinary approach and international collaborations wherever possible, is absolutely essential. Various member states in the Joint Study have provided valuable inputs for innovations in fast reactors and associated fuel cycles. A beginning has also been made in agreeing on areas of collaboration in R&D. This situation needs further discussion to clearly define topics of R&D and to work towards viable approaches which could be bilaterally or multilaterally pursued. It is clear that R&D is essential in all the areas defined by INPRO, namely, economics, safety, waste management, infrastructure, environment, security and proliferation resistance. This report describes those areas of R&D which emerged from the assessment of CNFC−FR using the INPRO methodology, and where there is a clear consensus among the States participating in the Joint Study on the programmes to be pursued. The R&D and collaborations discussed in this chapter mainly deal with sodium cooled fast reactors and associated fuel cycles, based on aqueous reprocessing integrated with waste management. The sodium cooled reactor emerged as a consensus system among the participants of the Joint Study due to early maturity and commercial exploitation. However, the Joint Study also recognized that several other fast reactor systems, e.g., a gas cooled and a heavy metal (Pb−Bi, Pb) cooled fast reactor are being considered / developed by member states. While R&D for these reactor systems would be similar in some of the generic aspects, it is realized that they would be quite different with respect to some other aspects. The Generation-IV programme also includes these reactor systems as candidates for development, and thus R&D on these systems is being pursued by interested member states. There is a large variation in the R&D facilities and expertise available in countries who participated in the Joint study; this is especially true for fuel cycle related programmes. In the current stage of discussions and assessments, input from participants has been largely in the area of economics and safety aspects of reactors and only a few inputs have been received with respect to the fuel cycle; there is also interest for pursuing R&D in a collaborative way, to further enrich and enhance the INPRO initiative.
12.2. ECONOMICS
A closed nuclear fuel cycle with fast reactor has to be economically attractive in order to gain more political and public acceptance. This implies that capital costs and operating costs of the fast reactor as well as fuel cycle facilities have to be addressed in a comprehensive manner. Economy can be achieved by simplification of designs, equipment and operating procedures besides introducing a high level of on-line instrumentation and monitoring as well as innovative inspection, repair and remote handling technologies. The innovations required for meeting these directions are discussed in various chapters of the report dealing with the individual subjects of
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economy particularly in Section 6.5. However, this section summarizes specific areas of R&D that have to be addressed. The capital cost of construction of fast reactors can be decreased by reducing the reactor size (per MW(e) generating capacity), shortening the construction time and by reducing the number of components in the reactor and their size. For example, the sodium cooled fast reactor proposed by JAEA is expected to have an electricity output of five times that of the prototype reactor MONJU, with a site area approximately only a ¼ of the area needed for MONJU. Such a large reduction in the reactor size is a major factor to reduce the capital cost significantly because the main components of the reactor costs (per MW(e)) are related to the total quantity of concrete and steel used in its construction. In fact a French analysis has established that the main component of costs of the electricity production by a sodium cooled fast reactor that needs to be addressed are the capital investment costs. The measures required for such a reduction in size include paradigm shifts in the design of reactor components such as use of a compact main vessel without cooling, integration of all core cover components, in-vessel sodium purification and a rationalized seismic design approach. As said above, the unit energy cost of electricity production can be reduced by shortening construction time as well as by increasing the service life of the reactor. This requires R&D on advanced core structural materials, manufacturing technologies, and in-service inspection technologies especially for under-sodium inspection inside the reactor vessel. Research on materials behaviour in a wide range of steady state as well as transient conditions, especially, under the presence of high levels of radiation, would provide confidence and requisite breakthroughs for enhancing the life of components. A comprehensive international effort on the design of components and buildings, including seismic designs, can lead to practically achievable and cost effective solutions that can simultaneously meet the safety requirements for reactor operation. The impact of R&D on reduction of capital cost of construction of fast reactors is impressively brought out in Table 12.1 and Fig. 12.1 based on data from the Russian Federation. TABLE 12.1. COMPARISON OF SPECIFIC CHARACTERISTICS OF FAST REACTORS RESULTING FROM R&D
It can be seen from Table 12.1 and Fig. 12.1 that building volume, construction area, amount of reinforced concrete used, amount of steel and other metals used per unit energy generation will
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be brought down considerably in the case of a BN-1800 as compared to a BN-800. Similarly, in the case of the European fast reactor (EFR), specific weight steel in tons per kW electrical has been brought down by innovations in design as compared to SUPERPHENIX (Fig. 12.2). Thus, one can see the commonality and confidence in pursuing this approach. The availability of advanced irradiation tools is an important constraint in the development of advanced structural materials for fast reactors. Projects such as the JHR reactor in Cadarache, France, will provide irradiation facilities with a high level neutron flux and enhance instrumentation that would be extremely important in characterizing nuclear materials.
BN-1800 BN GT-300 BN-350 BN-600 BN-800 BN-600M BN-900 5000 50
40
4000 30 Specific metal content of reactor,t/MW(e) (design) 20
3000 13.7
10.5 10 8.2 9 8 2000 7 5.3 6 1900 4.5 5
4 1500 3 3
1217 2 Specific capital cost, $/kW (design) $/kW cost, capital Specific
1000 1050 900 920 890 1 800 0.9 0.8 700 0.7 700 0.6 600 0.5 0.9 2345 1 FIG. 12.1. Trends in specific capital cost reduction due to investment in R&D: cumulative investments into R&D vs. R&D investments of first design (BN-350).
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Normalized steel specific weight t/kWe
Comparaison des masses par kWh de EFR et SPX
Manutention 0.47 secondaire
Circuits d'évacuation de la puissance 1.01 résiduelle
Pompes secondaires 0.24
Circuits secondaires 0.35
Générateurs de 0.61 vapeur
Circuit primaire 0.36
FIG. 12.2. From SUPERPHENIX to EFR: comparison of steel consumption.
Accelerators are invaluable tools for studying radiation damage in materials; accelerator testing with dual beams combined with multi-scale modelling of materials behaviour in reactor environment, can yield important data that could be used for screening materials for reactor applications as well as for understanding the mechanisms of damage. Thus, an important element of international collaboration is building of such test facilities and their availability for utilization by all member states. In addition, advances in physical metallurgy and characterisation techniques can be exploited to develop alloys with a high performance for use in next generation fast reactors.
12.3. SAFETY AND ENVIRONMENTAL IMPACT
A safe operation of nuclear reactors and fuel cycle facilities is important to achieve public and institutional support for development of nuclear energy. The safe operation of the reactor also contributes to the economics of nuclear power, by enhancing the availability of facilities and reducing their maintenance costs. The Joint Study identified a number of areas where R&D can contribute significantly to an enhancement of safety as well as environmental impact of fast reactors and associated fuel cycles. Development of efficient and cost-effective shielding materials such as boride/rare earth combinations, source reduction by use of hard faced materials which do not get activated and in- vessel purification of primary sodium are important means that could contribute to reactor safety and reduce man rem exposure. Reduction of public exposure can be achieved by development of advanced cover gas purification technologies that would provide a high decontamination factor from the radioactive gaseous fission products. Advanced designs of the roof slab, providing near complete leak tightness for cover gas can reduce public exposure. Innovative designs of the condenser structure need to be developed for minimizing the temperature of condenser water discharged into water body.
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The French analysis has emphasised that realizing improvements in safety aspects of sodium coolant fast reactor designs requires a large R&D effort on core physics and technology design as well as on preventive surveillance and in-service inspection. Prevention of core accidents can be brought about by the development of automatic negative reactivity insertion devices, sodium pumps with high inertia and advanced cladding and wrapper materials. Re-criticality prevention after core melting is considered as an important area for enhancing safety and also reducing capital cost of fast reactors.
12.4. ECONOMY AND SAFETY OF FUEL CYCLE FACILITIES
As in the case of reactors, economy as well as safety of fuel cycle facilities has similar goals, which can be addressed by a robust design and operational practices. Fabrication technologies for example should involve robust techniques with lesser re-work on off-spec products. Since reprocessing and waste management facilities handle large quantities of irradiated fissile materials, safety of these fuel cycle facilities is important in development of a robust fast reactor fuel cycle. As compared to nuclear reactors, safety of fuel cycle facilities is more focused on issues such as criticality, corrosion of key equipment such as dissolver, evaporator and waste tanks, and on processes for efficient recovery of radioactive materials. Increasing robustness and availability of crucial process equipment in reprocessing as well as waste management facilities is key to enhancing the economics of fuel cycle operations. The size of hot cells housing processing equipment is governed by the need to ensure sub-criticality in all conditions of operation. Development of materials and equipment which can reduce the size of process equipment as well as the space to be provided between equipment, without sacrificing criticality margins, would contribute to economy. The life of a re-processing and waste management plant is mainly governed by the life of the in-cell equipment, which faces hostile conditions of concentrated nitric acid and high levels of radiation. Development of materials for dissolvers, evaporators, etc., attempts to reduce corrosion loss thereby enhancing the life of this equipment. It is also important to adopt a robust welding technology as well as provide for on- line in-service inspection in order to ensure low frequency of leaks due to failure of pipe lines or welding. Thus, advanced condition monitoring systems that would provide inputs regarding health of the equipment is a crucial area of R&D for fuel cycle facilities. With respect to flow sheets used for reprocessing and waste management, the R&D is targeted towards ensuring recovery of uranium and plutonium without mutual separation, and towards reprocesses for recovery of long lived minor actinides and fission products to simplify waste management and reduce demand for repository space, besides resulting in lower exposure to personnel. Separation processes have to be designed to avoid or reduce use of reagents that add to the salt content of the waste. Also separation of industrially useful fission products such as caesium and palladium will result in societal benefits; this is an attractive feature of fast reactors due to the use of Pu rich fuels and the high burn-up achieved. The development of ceramic matrices with long term stability and higher capacity for waste loading is another important area of R&D which could enhance public acceptance of nuclear fuel cycles.
12.5. INTERNATIONAL COLLABORATION
The Joint Study has brought out a number of areas for improvement in the design and operational philosophies that could make fast reactors with a closed fuel cycle more economic, safe and
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proliferation resistant through innovations. It is obvious that the work involved is large and it is logical that collaborations would enrich R&D and compress the time frame for development. Also, comparison and benchmarking of results obtained by more than one country is a desired route for enhancing confidence in the approaches. These factors underline the importance of collaborative R&D in selected areas of reactor and fuel cycle technologies. From the national scenarios provided by the participants of the Joint Study, it is clear that the areas of emphasis of different member states could be different. For instance, some of the participants are interested in development of metallic fuel based reactor systems, while others would continue the emphasis on oxide, carbide or nitride fuels. The reference CNFC-FR has tried to address this issue and has chosen for the assessment a system based on consensus. However, the Joint Study has clearly identified options for bilateral or multilateral collaborations between countries having similar interests in spite of the fact that it may not represent a consensus of all the countries. Also, it is possible to identify programmes for international collaboration in generic areas such as materials development and testing, in-service inspection technologies, modelling and code development, methodologies for probabilistic safety assessment of fuel cycle facilities and harmonized regulatory approaches, to name a few examples. In the framework of the second phase of INPRO, the following collaborative projects related to fast reactors have been proposed:
• Integrated Approach for the design of safety grade decay heat removal system for liquid metal cooled reactors (DHR) (proposed by India); • Assessment using the INPRO methodology on the basis of general characteristics and input/output parameters of advanced and innovative nuclear fuel cycles within a large scale INS based on a CNFC concept to satisfy principles of sustainability in the 21st century (FINITE) (proposed by the Russian Federation); • Investigation of technological challenges related to the removal of heat by liquid metal and molten salt coolants from reactor cores operating at high temperatures (COOL) (proposed by India); • Proliferation resistance analysis and methodology improvements for multinational INS (PROMM) (proposed by France); • A global architecture of INS based on thermal and fast reactors including a closed fuel cycle (GAINS) (proposed by the Russian Federation taking into account results and recommendation of the Joint Study). It is expected that in the next phase of the INPRO project, more such activities will be conceived and implemented. It would be fruitful to consider the sharing of costly and state-of-art facilities among the countries developing fast reactors with a closed fuel cycle, to realize faster exploitation of CNFC-FR with innovative parameters.
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[48] OECD, NUCLEAR ENERGY AGENCY, Latest trends in nuclear fuel cycles- economic, environmental and social aspects, OECD, NEA, Vienna (2001). [49] MARTIN, PH., et al., French Program towards an Innovative Sodium Cooled Fast Reactor, Nuclear Eng. and Technology 39 (4), August (2007). [50] JAMES, N.J., SHEPPARD, G.T., “Red-oil Hazards in Nuclear Fuel Reprocessing”, Nuclear Engineering and Design 130 (1991) 59. [51] RAJ, B., KAMACHI MUDALI, U., Materials development and corrosion problems in nuclear fuel reprocessing plants, Progress in Nuclear Energy 48 (4) 283, May 2006. [52] INTERNATIONAL ATOMIC ENERGY AGENCY, Guidance for the application of an assessment methodology for innovative nuclear energy systems, INPRO Manual – Safety of Nuclear Fuel Cycle Facilities, Volume 9, IAEA-TECDOC-1575 Rev 1, Vienna (2009). [53] OECD NUCLEAR ENERGY AGENCY, Critical Operation Actions – Human Reliability Modelling and Data Issues, NEA/CSNI/R(98) 1, OECD, Paris (1998). [54] CARRÉ, F., WARIN, D., DELPECH, M., BROSSARD, PH., Global actinides management scenarios: transition from Generation-II / Generation-III to Generation-IV systems – Case of the French fleet, Proc. Of Int. Conf. (ICAPP 2005), May 2005, Seoul, Republic of Korea. [55] VASILE, A., FIORINI, G.L., CAZALET, J., LINET, F.L., MOULIN, V., GRENÈCHE, D., Assessment of a French scenario with the INPRO methodology, Proceedings of Int. Conf. (ICAPP ’06), Reno NV, USA, June 4-8, 2006. [56] INTERNATIONAL ATOMIC ENERGY AGENCY, Department of Safeguards, proliferation resistance fundamentals for future nuclear energy systems, IAEA-STR-332, Vienna (2002). [57] INTERNATIONAL ATOMIC ENERGY AGENCY, Guidance for the Application of an Assessment Methodology for Innovative Nuclear Energy Systems, INPRO Manual – Proliferation Resistance, Volume 5, IAEA-TECDOC-1575, Vienna (2007). [58] CAZALET, J., Future nuclear cycle systems and technology enhancing proliferation resistance, The International Nuclear Non-proliferation Science and Technology Forum, Tokyo, Japan, May, (2006). [59] HAHN, D.H., KIM, Y.I., KIM, S.O., LEE, J.H. and LEE, Y.B., Design Concept of KAIMER-600, Proceeding of Int. Conf. (GOLBAL 2005), Tsukuba, Japan, Oct., (2005). [60] INTERNATIONAL ATOMIC ENERGY AGENCY, Guidance for the Application of an Assessment Methodology for Innovative Nuclear Energy Systems, INPRO Manual – Infrastructure, Volume 3, IAEA-TECDOC-1575, Vienna (2007). [61] DELPECH, M., et al., “Global actinides management scenarios: transition from Gen II / Gen III to Gen IV systems – Case of the French fleet” Proc. of Int. Conf. (ICAPP 2005), Seoul, Republic of Korea, May 15-19, 2005. [62] BOUCHER, L., et al., “COSI: a simulating software for a pool of reactors and fuel cycle plants” Proc. of 13th Int. Conf. on Nuclear Engineering (ICONE 13), paper 50253, Beijing, China, May 16-20, (2005). [63] GARZENNE, C., et al, “Assessing Environmental and Health Impact of the Nuclear Fuel Cycle: Methodology and Application to Prospective Actinides Recycling Options” Proceedings of Int. Conf. (GLOBAL 2005), Tsukuba, Japan, Oct 9-13, 2005.
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ANNEX A OUTLINE OF A FRENCH CASE STUDY
The French case study on assessment of a national nuclear scenario with the INPRO methodology was completed in 2006. It concerned the application of the INPRO methodology to a French scenario, i.e. on the transition from present LWRs to EPRs in a first phase and to 4th generation fast reactors in a second phase during the 21st century. The scenario also considered the renewal of present fuel cycle facilities by third and the fourth generation ones. A detailed report on the results of this national assessment was presented to the STC of the Joint Study as a part of the French contribution to the project. The outline below presents main results and conclusions of the French case study. A1. INTRODUCTION One of the INPRO project main objectives is to develop a methodology for the assessment of innovative nuclear reactors and fuel cycles (INS) at a national, regional and global level, with respect to a number of important performance characteristics in the areas of economics, safety, environment, waste management, proliferation resistance and infrastructure. This assessment process ultimately should provide information about the potential sustainability of any INS evaluated. INPRO in its initial Phase, called Phase 1A, has developed a set of basic principles, user requirements, indicators, and acceptance limits in different areas [5]. In the following phase, called Phase 1B (first part) this set of requirements was tested by several national case studies, and studies performed by individual experts. The methodology was then updated [2]. During the ongoing phase of INPRO, called Phase 1B (second part), this methodology is applied to assess several complete INS. Several case studies are implemented by INPRO Member States with the objective to assess national INS and to identify the role and potential of innovative nuclear energy systems in satisfying the conditions for sustainable development and contribution to the solution of energy problems at a global, regional and national level during the 21st century. The French contribution to this work does address the application of the INPRO methodology to the “French case”. The assessed scenario covers the transition from the present LWRs fleet to one with EPRs in a first phase and to Generation-IV fast reactors deployment in a second phase, during the 21st century [61]. The scenario also considers the renewal of present fuel cycle facilities during this century. A2. THE CURRENT FRENCH NUCLEAR FLEET The present French nuclear fleet is composed of 58 PWRs with a total capacity of about 60 GW(e) and a total electricity production of about 400 TWh per year. The total cumulated operational experience is about 1200 reactor years (Fig. 3.2 of Chapter 3). Most of the reactors were started up during a short period of less than twenty years, and, even if their lifetime will be extended at least for some of them beyond 40 years, their partial replacement by EPRs (European pressurized reactor) will start around 2020. Concerning fuel cycle facilities, the present situation and their expected evolution are presented in the next section.
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A3. THE TRANSITION SCENARIO A3.1. Description of the reference scenario assessed and hypothesis The starting point of the scenario is the year 2010, and it is supposed that the energy supplied by a 60 GW(e) reactor fleet is constant and equal to 400 TWh(e) per year. About 91 % of this energy is produced by UOX fuels and 9 % by MOX fuels at 45 GWd/t mean burnup. By the year 2010 the total plutonium inventory is 250 tons with • 180 tons stored in spent fuels, o 115 tons in UOX fuels; o 60 tons in MOX; o 5 tons in spent URT fuels. • 20 tons in fuel cycle plants • 50 tons in reactor cores.
Average burnup of UOX and MOX fuels are increased to 60 GWd/t by 2013.
Until 2020, the UOX spent fuels processing addresses the partitioning of plutonium alone, under “just-in-time” Pu flow conditions (processing, manufacturing, reactor insertion) are extended, thus not all spent fuel assemblies are processed. The produced glasses contain fission products, MA, and Pu processing losses (0.1 %). In a later period, around 2040, Generation-IV fast neutron systems, designed for a grouped recycling of all actinides appear as the best candidate nuclear systems. They will efficiently use natural uranium and will be environmentally friendly thus contributing, among other energy sources, to sustainable development. This leads to develop interesting scenarios benefiting from industrial plant renewals (reactors and fuel cycle plants), while optimising the utilization of existing plants and taking benefit from their replacement to implement innovative processes and technologies. The reference scenario considers grouped recycling of all actinides (U-Pu-MA). It foresees around 2040 both the deployment of fast reactors in the French generating fleet (sodium or gas cooled), and the start-up of a new spent fuel treatment plant, superseding the existing plant of La Hague. The effective deployment of this scenario is linked to a progress sought on the quality of ultimate waste forms, and the technical and economical optimisation of a geological repository (the design of which depends on the long term decay heat). Main trends for reactors and associated fuel cycles are summarized in Fig. 3.7 of Chapter 3. For simplification and modelling purposes, the reference scenario for the transition phase is considered as follows: • 2020–2035 Start of renewal of 50 % of the fleet with EPR reactors; this renewal relates to the end of the service life of the first PWR plants introduced in 1975/1985 and is carried out, depending on EDF prospects, at the rate of 2 GW(e) per year. • 2035–2040 Start of renewal of the remaining 50 % reactors of the previous generation by 4th generation fast neutron systems. Implementation of the advanced processing of spent MOX fuel for plutonium and minor actinides recycling in 4th generation fast neutron systems.
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• 2080 Start of replacement of the EPRs which were first introduced in 2020 by 4th generation fast reactors. The following deployment scheme (Fig. 3.8) was modelled by the COSI code [62] for the period 2000–2100. Additional assumptions: • Constant 60 GW(e) installed capacity producing 400 TWh(e)/year during the period; • The order in which existing PWRs are decommissioned is CP0, MOXed CPY, unMOXed CPY, P4, P4’and finally N4; • The minimum spent fuel (SF) cooling time before processing are: o 5 years for PWRs SF; o 2 years for 4th generation fast neutron systems, which are considered in the study to be self-breeders or slight breeders. • Two types of fast reactors were considered in this study: o SFR (sodium fast reactor); o GFR (gas cooled fast reactor); In both cases closed fuel cycle with global actinides recycling is assumed. • Advanced partitioning or grouped partitioning is started in 2020, at the same time as: o Interim storage of minor actinides; o Production of so-called « low thermal and low radiotoxicity» glass, for storage in a small area (volume); o Processing of spent UOX fuel; o Processing of spent MOX fuel if necessary. • From 2020 to 2025, plutonium is recycled using single plutonium recycling in PWRs; from 2025 to 2035, UOX and MOX spent fuels are put into interim storage and processed; plutonium and minor actinides are recycled in GFR fuels; • From 2035 to 2050: GFRs uniformly recycle all actinides and resorb the stock and production of the previously existing 2nd generation PWR fleet in 15 years; • In 2080: Second phase in renewal of the reactor fleet, since the firstly implemented EPRs in 2020 will have reached the end of their 60-year life and will be replaced by GFRs.
A3.2. Results of the scenario’s analysis in terms of materials flows and inventories
Results obtained for this reference scenario are summarized in Table A1.
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TABLE A1. MASS INVENTORIES (TONS) IN THE FUEL CYCLE.
2035 2050 2070 2100 Natural U (annual, t ) 7850 4200 4200 0 Natural U (cumulated6, kt) 430 515 600 660 SWU (annual, M SWU/a) 6 3.2 3.2 0 Pu (Total, t) 7 455 576 685 848 Np, (t) 22 28 30 22 Am, (t) 50 61 65 47 Cm, (t) 4 7 10 17 MA (Total, t) 76 96 105 86 TRU total (t) 531 672 790 934 Fuel with TRU in fleet (%)8 0 50 50 100 TRU in storage, (t)9 27 28 29 30
A4. THE ASSESSMENT WITH THE INPRO METHODOLOGY In each of the six INPRO areas, economics, safety, environment, waste management, proliferation resistance and infrastructure, different basic principles are defined and accompanied by user requirements with their corresponding criteria. Each criterion includes an acceptance limit and an indicator. Almost one hundred criteria are defined to assess a nuclear energy system. We present here the main trends of the assessment under way on the previously presented French scenario.
A4.1. Economics
The cost of energy from innovative nuclear energy systems, in our case EPR in the first phase of the scenario and fast reactors in the second phase, taking all relevant costs and credits into account, is competitive with that of alternative energy sources available.
6 Cumulative needs for natural uranium since 1980, in tonnes. World reserves are conventionally accepted at 4 million tonnes, the world resources are estimated at 11.4 Mt (<130 $/kg) (cf. Red Book of the AIEA/OECD).
7 Total = Temporary storages, Reactors, Factories, Disposal – These values include the Pu from the decay of Cm in temporary storage. Pu production by the stored Cm later recovered is high above all for alternatives 2 (20 tonnes in 2100) and 3 (32 tonnes in 2100). 8 Reactor fraction in the fleet containing actinides (Pu or minor actinides). 9 Inventory of TRU in tonnes (Pu, Np, Am and Cm) in ultimate storage (vitrified waste), in long-term storage of non-recyclable irradiated fuels according to the scenario or specific materials (for instance: Cm in the case of alternative 3). The loss rate during spent fuel processing is 0.1 % for all actinides.
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The EPR offers several advantages: • Shorter construction time (57 months) and lower construction costs; • Longer service life of 60 years; • Increased availability, reaching more than 90 % over the entire service life of the plant; • Lower operating costs; • Optimised use of fuel resulting in 15 % savings on uranium. This means that a reactor series of this type would generate base-load electricity at a cost of almost 20 % less than that of the most advanced high-efficiency, high-output combined cycle gas power stations in France. For the EPR, the total cost is assessed considering the investment (17.0), the operation and maintenance (5.8), the fuel (4.2 €/MWh), dismantling (0.1 €/MWh) and the R&D (0.6 €/MWh). The total cost is 27.7 €/MWh. The same for the Gas Combined Cycle (GCC) gives 33.9 €/MWh. The cost of the electricity produced by the EPR is around 10 to 20 % less than the one for the GCC. The cost for the EPR does consider the externalities not taken into account for the GCC. For the GFR/SFR the objective is to achieve a design the cost of which has to be comparable or less that the one of the innovative LWR (EPR). Obviously the uncertainties are larger. Concerning the total investment required to design, construct, and commission innovative nuclear energy systems, including interest during construction, it should be such that the necessary investment funds can be raised. Among the relevant figures of merit, the total investment required should be compatible with the ability to raise capital in a given market climate. The comparison focused on the capital cost between a nuclear system and a fossil conventional system is systematically unfavourable to the nuclear. The competitiveness is found on the cost of the KWh. With regard to the current situation, it is reasonable to consider that this established fact should improve by means of the reduction of the durations of construction. This duration is estimated in 57 months for the EPR and it is aimed 48 months for the Generation IV concepts. This situation is acceptable for a developed country like France and there are no economic showstoppers for the deployment of nuclear systems. An important aspect in the economic area is the risk of investment in innovative nuclear energy systems; it should be acceptable to investors taking into account the risk of investment in other energy projects. Considering the past large experience gained in France, no major issues are expected to license both the EPR and the innovative Generation-IV systems of GFR or SFR. The rationale of the statement is that even if an innovative safety approach is considered for the Generation IV systems the foundation of this approach remains conventional and coherent with the current key principles. Moreover an effort is planned within the Generation IV context to improve the degree of harmonisation for the national licensing constraints. The project construction and commissioning times should be comparable with those for other energy supply alternatives. As indicated above, the construction time expected for the EPR is 57 months. The experience gained through the construction of the French fleet lead to consider the figure as realistic. Uncertainties are still present for Generation IV systems but the improvement of the construction techniques within the next 30 years can contribute to consider that the objectives for the construction time (48 months) will be met. Concerning the political environment, relevant indicators show long-term support for nuclear power. The flexibility of the INS is also considered in this area. Two sorts of “flexibility” have to
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be considered: the capability to serve different types of markets and the possibility to achieve different modes of energy production: base load or load following. The EPR being a large size plant can easily fit with markets/grids able to accept such large plants. Finland and France are very different in terms of global electricity production capability but the EPR can answer the expectation of these two markets. The EPR is able to answer the two function modes: base load or load following. Obviously, this will be done with different performances in terms of maximum burn up: 57 GWd/t for the base load and 52 GWd/t for the load following mode. Analogous requirements will be answered by the Generation-IV systems. Moreover the range of Generation-IV plant sizes is larger (300–1200 MW(e)) opening the possibility for wide market applications. With a strategy of fuel recycling, the current installations for the fuel cycle do not present the same flexibility. Large plants as La Hague have to be considered for large national markets (as the French one) or on a regional basis. With the Generation-IV systems the situation can evolve toward a greater flexibility. The Generation-IV requirements for fast reactor systems ask for onsite fuel cycle installations and an R&D effort is implemented to look for the compactness of the process and installations.
A4.2. Safety
The development of the EPR safety approach was founded by a joint safety approach of French and German safety authorities for future PWRs in parallel with the development of the EPR design. The initial safety goals were the following: Safety improvement compared to existing PWRs; design to be licensable in France and Germany; evolutionary design; harmonisation between two countries’ approaches. Among the important EPR’s general safety objectives it seems important to recall: A further reduction of the core melt frequency; the “practical elimination” of accident situations which could lead to large early releases of radioactive material; for low pressure core melt situations the design has to be such that the associated maximum conceivable releases would necessitate only very limited protective measures in area and time; no permanent relocation and no need for emergency evacuation outside the immediate vicinity of the plant; limited sheltering and no long- term restriction in food consumption. The approach for the definition of the safety related architecture relies on the following safety principles: Defence-in-depth remains the fundamental principle but must be extended with an independence between the different levels of defence; design should be made on a deterministic basis, supplemented by probabilistic methods; utilization of operating experience and results of in-depth safety studies in France and Germany; utilization of results of research and development work. The greater independence between the different levels of defence is achieved through: diverse systems for each level; highly reliable support systems; diesel generators; containment heat removal cooling chain; spatial separation of redundant trains as a measure against internal hazards (fire, flooding). Moreover, operating and safety functions are separated to simplify system layout; together, the operating and safety provisions provide progressive responses commensurate with any abnormal occurrences. One of the EPR’s main features is its simple design based on the 4 primary loops and 4 safety trains concept which applies to mechanical
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equipment as well as the electrical power supply and the associated instrumentation and control (I&C). The achievement of the above goals and objective as well as the implementation of the above principles enable the EPR to meet the highest level of safety required worldwide for future nuclear power plants. For the Generation-IV systems the details for the safety approach are still under discussion. The basic idea – from French side – is to try improving the safety achieved with the EPR both in terms of “level” of safety and in terms of “quality” of safety. Obviously, the process of certification for these systems has to fit with a context which, while remaining coherent with the current practices, will integrate new tools such as the simultaneous use of the determinist and probability techniques through, for example, a risk informed approach. This last one could become an integral part of the design/evaluation approach which, generally speaking, will have to answer the following criteria: be in agreement with the current safety standards and those that can be foreseen for the future; prove full agreement with the principles of the defence in depth; allow for the design/evaluation stages, to manage simultaneously deterministic and probabilistic criteria; address threats and hazards so as to guarantee the consideration of internal initiating events within the framework of an homogeneous approach; improve the demonstration capabilities in the domains where gaps still exist in the current state of art; demonstrate a reduction of the risk level with regard to current systems. Such an approach has to allow optimizing the safety architecture in terms of performances, reliability and costs, notably by a better definition of the requirements, and the classification of the component of this architecture. In conclusion, in regard to the safety of future French fleet systems, it is thus reasonable to expect a global improvement of the answers to the criteria identified by the INPRO methodology. A4.3. Environment The impact on the environment of the near future systems based on EPR technology and of later systems based on fast reactors is, by design, in an acceptable range for national, European and international standards. Concerning the use of fissile materials, the replacement of current and Generation II PWRs by EPR will improve the natural uranium consumption by about 15 %. Three main characteristics of the reactor to achieve this figure are: • Better use of the fuel due to the larger core compared to current cores; • Neutron leakage reduced by a heavy reflector at the periphery of the core; • Improved mean unloading burn-up of 60 GW/t(HM). In the second phase of the scenario, fast reactors allow dramatic savings in natural resources during the century [63], as indicated in Table A1. A4.4. Waste management In the INPRO waste management area, the first basic principle concerns the minimization of waste production. For long lived waste this is achievable with fast reactors due to their better efficiency compared to thermal reactors and the specific characteristics of the neutron spectrum. By design, future fast reactors will minimize the generation of waste at all stages.
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The second waste management basic principle is related to the protection of human health and of the environment: radioactive waste in an INS shall be managed in such a way as to secure an acceptable level of protection for human health and the environment, regardless of the time or place at which impacts may occur. The third waste management basic principle deals with the burden on future generations. In the foreseen scenario this burden is minimized by recycling all actinides and then obtaining a reduction in waste masses to the final repository. The last basic principle of the waste management area is related to the interactions and relationships among all waste generation and management steps and their optimization. This is accomplished by an adequate classification of the radioactive waste arising from the INS. This scheme permits unambiguous, practical segregation and measurement of waste arisings. Intermediate steps between generation of the waste and the end state should be taken as early as reasonably practicable. The design of the steps should ensure that all-important technical issues (e.g., heat removal, criticality control, confinement of radioactive material) are addressed. The processes should not inhibit or complicate the achievement of the end state. For present PWRs the overall waste management scheme was not fully addressed during the design stage. Potential for optimization of effluents management are implemented for EPR. All these issues are addressed by present work done in the frame of the French 1991 Act. A complete chain of processes from generation to final end state is sufficiently detailed to make evident the feasibility of all steps. All fuel cycle steps are addressed by design in future INS. A4.5. Proliferation resistance The first basic principle on proliferation resistance deals with features and measures implemented throughout the full life cycle for innovative nuclear energy systems to help ensure that INSs will continue to be an unattractive means to acquire fissile material for a nuclear weapons programme. In this framework, the commitments, obligations and policies are regarded as acceptable by the international community. The attractiveness for a nuclear weapons programme of nuclear material in this INS deployment scenario is considered as a key criterion. In this respect the main characteristics of the fuel cycle must be underlined: • Fast reactor cores considered have a breeding ratio equal to one. No radial blankets are used to produce plutonium; • Closed cycle avoids “plutonium mines” in repositories; • All actinides, uranium, plutonium and minor actinides are kept and recycled together so there is no separated pure plutonium in any step of the fuel cycle; • Present light water reactors are progressively replaced by fast reactors so there will be no need of enriched uranium and enrichment technologies and plants. The use of fast reactors with onsite treatment and fabrication with grouped actinide management makes the diversion of nuclear material reasonably difficult and detectable. The second proliferation resistance basic principle states that both intrinsic features and extrinsic measures are essential, and neither shall be considered sufficient by itself. Innovative nuclear
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energy systems should incorporate multiple proliferation resistance features and measures. Each acquisition path must be covered by appropriate verification measures and covered by robust barriers. The combination of intrinsic features and extrinsic measures, compatible with other design considerations, should be optimized (in the design/engineering phase) to provide cost- efficient proliferation resistance. Intrinsic features for proliferation resistance are included at the design stage of INS (and associated fuel cycle facilities) alleviating the non-proliferation costs. An important feature of this concept is the drastic reduction of transportations of nuclear material between several sites. The verification approach with a level of extrinsic measures was agreed to between the verification authority (e.g., IAEA, regional safeguards organizations, etc.) and responsible national organizations.
A4.6 Infrastructure
The INPRO infrastructure basic principle states that regional and international arrangements shall provide options that enable any country that so wishes to adopt INS for the supply of energy and related products without making an excessive investment in national infrastructure.
In our case, a national legal framework is already established covering the issues of nuclear liability, safety and radiation protection, control of operation and security, and proliferation resistance by the Basic law of 1963 plus basic safety rules.
Safety and radiation protection arrangements are also established in the European framework and their standards apply to the French nuclear system. Concerning the industrial and economic infrastructure of the country, it is supposed to be adequate to support the project during construction and operation. The size of installations to be deployed matches the local needs and support infrastructures are internally available.
In the field of socio-political infrastructures, adequate measures are taken to achieve public acceptance of the planned INS installation. The information to public is coordinated by the CSSIN (high council for nuclear safety public information), and the participation of public in decision-making process (to foster public acceptance) is implemented by the CLI (local information commission) and the CPDP (particular commission for the public debate). Long term commitment is considered as sufficient and established by the French national act on energy supply and the Atomic Energy Committee.
Finally, public acceptance of nuclear power is sufficient to ensure there is negligible political or policy risk to the investment. Instruments are implemented to check continuously the public acceptance. e.g., in France a periodic standardized poll is implemented under the responsibility of safety authorities.
In addition, the necessary human resources are available to enable the operating organization to maintain a safety culture to achieve safe operation of the INS installation. The present operating organization, Electricité de France (EDF), has enough knowledge of the present plants to be an intelligent customer and should keep a stable framework of trained staff.
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To maintain high standards in safety culture, periodic safety review mechanisms are implemented, covering technical infrastructure and management areas. This includes a peer review from IAEA and periodic inspections and audits by the safety authority plus a periodic (10 years) in depth revision for licensing renewal.
A5. CONCLUSIONS
The INPRO methodology was used to assess a French reference scenario, on the transition from present LWRs to EPRs in a first phase and to 4th generation fast reactors in a second phase during the 21st century. Improvements were pointed out on the six areas considered: economics, safety, environment, waste management, proliferation resistance and infrastructure. The cost of the electricity produced by the EPR is around 10 to 20 % less than the one for the GCC. The cost for the EPR does consider the externalities not taken into account for the GCC. Comparisons focused on the capital cost between a nuclear system and a fossil conventional system is systematically unfavourable to the nuclear. Lower construction delays will increase attractiveness of future nuclear systems compared to other ones. For the GFR/SFR the objective is to achieve a design the cost of which has to be comparable or less than the one of the innovative LWR (EPR). Safety features of EPR design were underlined measures to prevent severe accidents and to minimize their consequences. Future fast reactors will be designed taking into account the experience gained from the past and present designs and taking advantage from the design and operation of the current reactors. The main characteristics of fuel cycles envisaged are coherent with the requirements for improved proliferation resistance. In the waste management area the transition from current PWRs to Generation-IV fast reactors allows significant gains in waste production and management. The whole infrastructure needed for the deployment of future nuclear systems is well established in the French context.
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ANNEX B JAPAN’S EVALUATION RESULTS OF THE INPRO JOINT STUDY
B1. INTRODUCTION Japan has launched a research and development project on a fast reactor cycle system named “Fast Reactor Cycle Technology Development” (hereinafter referred to as FaCT) in 2006. In this project, Japan is implementing research and development on Japan’s innovative fast reactor concept named “Japanese Sodium Fast Reactor” (hereinafter referred to as “JSFR”) and its fuel cycle systems. Therefore, we considered the JSFR as an INS in the INPRO methodology and evaluated it in terms of the following aspects: economics, environment, safety, waste management, proliferation resistance, infrastructure, and scenario analysis. The JSFR is a sodium cooled, MOX (or metal) fueled, advanced loop type reactor that evolved from Japan’s fast reactor technologies and experience (see also Chapter 3). There are two options for the reactor size: a large scale reactor generating 1500 MW(e) and a medium scale reactor generating 750 MW(e) with a similar plant configuration. The large-scale sodium-cooled reactor utilizes the advantage of “economy of scale” with electricity output set to 1500 MW(e); it is designed as a twin plant (total electricity output of 3000 MW(e)), and to reduce the amount of materials by adopting innovative technologies. Oxide fuel, which has been intensively developed, has sufficient potential to satisfy all design requirements and thus is selected as the reference fuel. Major specifications for the large-scale JSFR are shown in Table B.1. The JSFR is a design concept recognized as a promising candidate for Generation- IV systems.
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TABLE B.1. MAJOR SPECIFICATIONS OF JSFR DESIGN
Sodium-cooled reactor (1,500MWe) Design requirement MOX fuel (metallic fuel) Breeding core Break-even core Safety Out-of-pile and in-pile experiments are underway, concerning the passive safety mechanism and measures to avoid recriticality. Efficient utilization of Breeding ratio (1.0～approx. 1.2) 1.10 (1.11) 1.03 (1.03) nuclear fuel resources Fissile fuel inventory required for the initial loading core 5.7 (4.9) t/GWe 5.8 (5.1) t/GWe Time required to replace all nuclear power reactors with FRs Approx. 60 years － Reduction of Can accept MA content up to approx. 5% that is recycled from LWR spent MA burning environmental fuel under low decontamination condition (with FP content of 0.2 vol%). burden Having a possibility of transmuting self-generating LLFP (129I and 99Tc), by FP transmutation installing the FP both inside the core and the radial blanket region.
In-core average (150GWd/t or higher) 147 (149)GWd/t 150 (153)GWd/t
a) b) Whole-core average 90 (134)GWd/t 115 (153)GWd/t (60GWd/t or higher)
Economic Operation period 26 (22) 26 (22)
competitiveness c) (18 months or longer) months months Availability (Calculated value) (90% or more) Approx.95 (94)% Approx.95 (94)%
Reactor outlet temperature 550℃ d) Thermal efficiency / Onsite load factor 42.5% / 4%
e) Unit construction cost (¥200,000/kWe or less) Relative value：Approx. 90％
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B2. ECONOMICS JAEA (Japan Atomic Energy Agency) carried out a study on economics of the JSFR in the FaCT project. Japan has a program to introduce a sodium cooled type fast breeder reactor with MOX fuel (JSFR) around 2050. The JSFR is the reference type, with an electrical output of 1 500 MW(e), an average fuel burn-up including blanket fuels of 115 GWd/t(HM), a plant efficiency of 43 per cent, a load factor of 95 per cent, and a plant lifetime of 60 years. Main economic parameters are as follows: a discount rate of 2 per cent, 2005 money value, and an exchange rate of 110.16 Yen to the US dollar. In this economic study, a future LWR was selected as an alternative energy source instead of thermal power plants because this LWR is projected to be one of the cheapest power plants in the future. It was confirmed that the JSFR (NOAK plant) concept satisfied INPRO acceptance limits in all economic criteria and demonstrated competitive advantages as compared with the future LWR in most economic indicators. B2.1 Overview of evaluation results The detailed results of an evaluation of economics using the INPRO methodology are shown in Table B.2. In INPRO user requirement UR1.1 and UR1.2, Japan’s future LWR is expected to be the main competitor to fast reactors and in accordance with the INPRO methodology used for defining acceptance limits. The future LWR is characterized by an electrical output 1500 MW(e), a fuel burn-up of 60 GWd/t(HM), a plant efficiency of 35 per cent, a load factor of 90 per cent, and a plant lifetime 60 years. Regarding the evaluation of generation costs in the INPRO indicator IN1.1.1, it is found that the JSFR’s electricity production cost (CN: 15.10 mills/kWh) are cheaper than the ones from the future LWR (CA: 26.59 mills/kWh); therefore the JSFR has a good economic competitiveness in the future. The JSFR is assumed to be an nth of a kind (NOAK) type plant. The generation cost for the future LWR is obtained based on present ABWR and APWR. Also for IRR (internal rate of return) and ROI (return on investment) defined in IN1.2.1, the JSFR has economic advantages compared to a future LWR. As to IN1.2.2 “investment”, the construction cost including IDC (interest during construction) of the JSFR and a future LWR are 1615 and 2596 M$, respectively; therefore the JSFR is a more attractive option than a future LWR from the viewpoint of investment risk. In IN1.3.1 “licensing status” of UR1.3, Japan’s fast breeder reactor project is in the middle of the development stage and we are now addressing regulatory issues and costs included in the development proposal. For example, regarding the adoption of high burn-up fuels, data about high burn-up fuels from irradiation experiments are collected and its costs are included in the JSFR R&D costs. As to IN1.3.2 “reality of project construction and commissioning times”, a prototype reactor ‘MONJU’ has been built. The construction and commissioning schedules of the JSFR are set up based on the construction experience of ‘MONJU’. The construction time of the JSFR is assumed to be 7 years including design time because a sheet steel and concrete (SC) structure of the containment is adopted in the design of the JSFR in order to shorten construction time, which is almost the same as construction time of a future LWR. Regarding the evaluation of financial robustness in IN1.3.3, the robustness index (RI: CA/CN in IN1.1.1) of the JSFR is about 1.64. The JSFR is tolerant of future R&D and market uncertainties such as discount rate,
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construction time, and natural uranium price. Because there are mid and long term solid nuclear policies in Japan, IN1.3.4 “political environment” is stable. Typical Japanese policies are as follows: • “Framework for nuclear energy policy”, which is the basic nuclear policy amended every five years; • “New national energy strategy”, which claims that the ratio of nuclear energy to total power generation shall be maintained at about 30−40 per cent or more after 2030. Concerning IN1.4.1 “flexibility” of UR1.4, there is a possibility that some of the JSFR components and materials are able to be transferred to other markets. B2.2. Comments on INPRO methodology As for IN1.3.1 “licensing status” and IN1.3.4 “political environment”, since both indicators also have aspects of infrastructure, it may be necessary to overview and rearrange the indicators. TABLE B.2. RESULTS OF INPRO ECONOMICS CRITERIA EVALUATION
BP: Basic Principle BP1: Energy and related products and services from Innovative Nuclear Energy Systems shall be affordable and available. UR: User UR1.1: The cost of energy from innovative nuclear energy systems, Requirement taking all relevant costs and credits into account, CN, should be competitive with that of alternative energy sources, CA. that are available for a given application in the same time frame and geographic region. CR: Criteria Evaluation results IN: Indicator AL: Acceptance Limit IN1.1.1: Cost of AL1.1: CN < k*CA [Acceptance] nuclear energy, CN CN-JSFR 15.10 < 26.59 (CLWR) (mills/kWh) (CN-JSFR: The construction IN1.1.2: Cost of costs of the JSFR NOAK energy from plant are reference value.) alternative source, CA (CLWR : Power generation cost for CLWR is obtained for future LWR reducing power generation cost based on ABWR or APWR.) UR: User UR1.2: The total investment required to design, construct, and Requirement commission innovative nuclear energy systems, including interest during construction, should be such that the necessary investment funds can be raised. CR: Criteria Evaluation results IN: Indicator AL: Acceptance Limit IN1.2.1: Financial AL1.2.1: Figures of merit are [Acceptance] figures of merit comparable with or better than those for competing energy technologies of IRRJSFR 19 % ≧10 % comparable size. (%/year)
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AL1.2.1.1: Internal Return Rate IRRN > IRRLIMIT ROIJSFR 12 % ≧3 % AL1.2.1.2: Return of Investment (%/year) ROIN > ROILIMIT IN1.2.2: Total AL1.2.2: The total investments required [Acceptance] investment should be compatible with the ability to Investment JSFR 1615 M$≦ raise capital in a given marketing Investment LIMIT 2596 M$ climate. AL1.2.2: Investment N < Investment LIMIT UR: User UR1.3: The risk of investment in innovative nuclear energy systems Requirement should be acceptable to investors taking into account the risk of investment in other energy projects. CR: Criteria Evaluation results IN: Indicator AL: Acceptance Limit IN1.3.1: Licensing AL1.3.1: For deployment of first few [Acceptance] status NPPs in a country: Plants of same basic design have been constructed and Licensing JSFR operated. Plan to address regulatory For deployment of a FOAK plant in a issues available and costs country with experience operating included in development NPPs: Design is licensable in country of proposal. origin For development: Plan to address regulatory issues available and costs included in development proposal IN1.3.2: Evidence that AL1.3.2: For deployment of first few [Acceptance] project construction NPP in a country: Construction schedule - Construction schedule and commissioning times used in financial analyses have The prototype reactor times used in financial been met in previous constructions ‘MONJU’ has been built. analyses are realistic. projects for plants of the same basic Schedules analyzed to design. demonstrate that scheduled For deployment of a FOAK plant: A times are realistic taking into convincing argument exists that the account the experience with construction schedule is realistic and ‘MONJU’. consistent with experience with previous NPP construction projects carried out by Tct JSFR 7 ≦7 (year) the supplier and includes adequate (Tct JSFR includes design time contingency. for specific plant and For development: Schedules analyzed to construction time of 46 demonstrate that scheduled times are months.) realistic taking into account experience with previous NPP construction projects. IN1.3.3: Financial AL1.3.3: robustness index RI ≥1 [Acceptance] robustness index of - Robustness INS, RI. RI JSFR 1.64 > 1
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IN1.3.4:Long term AL1.3.4: Commitment sufficient to [Acceptance] commitment to enable a return on investment - Political environment nuclear option This is analyzed in the area of There is mid and long term Infrastructure – see text policy of nuclear option in Japan. UR: User UR1.4: Innovative energy systems should be compatible with the Requirement requirements of different markets. CR: Criteria Evaluation results IN: Indicator AL: Acceptance Limit IN1.4.1: Are the INS AL1.4.1Yes. - Flexibility components adaptable Note: The UR and Criterion are to different markets? considered only in brief in this manual.
B3. ENVIRONMENT In the Joint Study an evaluation of the environment area was carried out in Japan based on the INPRO methodology. Japan has been conducting the FaCT project to research and develop, and demonstrate a commercialized fast reactor cycle system in the future. A combined system of the JSFR with a MOX fuel core, an advanced aqueous reprocessing process named NEXT (new extraction system for TRU recovery), and a simplified pelletizing fuel fabrication was selected as the reference system in this study. The quantitative data for the reference system in Table B.1 were used in the evaluation as far as they were available. Otherwise data for existing LWR cycle systems or engineering judgments based on the characteristics of the system and experiences in the environment area were used in the evaluation. Furthermore, important data are published in the document of “OECD/NEA- IAEA, Uranium 2005: Resources, Production and Demand (2006)” by the following organizations: • IPCC: The Intergovernmental Panel on Climate Change; • IIASA: International Institute for Applied System Analysis; • WEC: World Energy Council.
As shown in the Chapter 8 “Scenario Analysis”, these data were also used in the evaluation.
B3.1. Evaluation results
The environment area of the INPRO methodology has two basic principles BP (BP1: acceptability of expected adverse environmental effects, BP2: fitness for purpose) and four user requirements UR. Each BP has two URs. The URs are as follows: • UR1.1: Controllability of environmental stressors. • UR1.2: Adverse effects ALARP (As low as reasonably practicable). • UR2.1: Consistency with resource availability. • UR2.2: Adequate net energy output.
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Table B.4 shows detailed evaluation results regarding the environment in the Japanese study. Concerning UR1.1, the important nuclides from the viewpoint of radiological effects on the public were evaluated in the FaCT project (Table B.3). The effective dose rates were calculated based on the radionuclides released from an INS at an assumed specific site. The INPRO criteria were fulfilled because the calculated dose rates were lower than the regulatory standards for effective dose rates. Almost all of the stressors mentioned in the INPRO manual will be evaluated based on the “environmental impact assessment (EIA) Law,” etc., when an INS will be introduced in Japan. Outline of the EIA law is shown in the Fig. B.1.
TABLE B.3. STRESSORS ON RADIO-NUCLIDES EVALUATED IN THE FACT PROJECT
Release per Average release Stressors unit of product per year Units Units Compartment Radionuclides 83mＫｒ 2.0×105 MBq/tHM 4.2×106MBq/a Air 85mＫｒ 8.4×105 MBq/tHM 1.8×107 MBq/a Air 85Ｋｒ 5.6×106 MBq/tHM 1.8×108 MBq/a Air 87Ｋｒ 3.0×105 MBq/tHM 6.3 ×106 MBq/a Air 88Ｋｒ 6.1×105 MBq/tHM 1.3 ×107 MBq/a Air 131mXe 2.1×107 MBq/tHM 4.5 ×108 MBq/a Air 133m 5 7 Reactor System Xe 9.8×10 MBq/tHM 2.1 ×10 MBq/a Air 133Xe 2.1×108 MBq/tHM 4.3 ×109 MBq/a Air (1500MWe) 135mXe 1.0×106 MBq/tHM 2.2 ×107 MBq/a Air 135Xe 1.1×107 MBq/tHM 2.3 ×108 MBq/a Air 138Xe 5.9×105 MBq/tHM 1.2 ×107 MBq/a Air 131I 3.1×101 MBq/tHM 6.4 ×102 MBq/a Air 133I 2.8 MBq/tHM 5.8 ×10 MBq/a Air 3H 3.7×107 MBq/tHM 7.3×109 MBq/a Air Fuel Cycle System 14 × 5 × 7 C 1.3 10 MBq/tHM 2.5 10 MBq/a Air (200tHM/y) 85Kr 3.8×108 MBq/tHM 7.6×1010 MBq/a Air 129I 3.2×103 MBq/tHM 6.4×105 MBq/a Air （Mainly Reprocessing）
Concerning UR1.2, the ALARA (as low as reasonably achievable) concept will be applied to the R&Ds in the FaCT. Concerning UR2.1, the evaluation results met the criteria CR2.1.1, CR2.1.2 and CR2.1.4, however the criteria such as CR2.1.3, CR2.2.5 and CR2.2.6 were not evaluated in the FaCT. Concerning UR2.2, it was not evaluated in the FaCT project. B.3.2. Comments to the INPRO methodology (1) The relationship to the “Environment Impact Assessment” laws and acts, which would have been established in almost all of the INPRO member countries based on the ”Basic Environmental Law” (1993) of the United Nations, should be mentioned in the manual because the assessment items are almost common between them as follows: • The environmental stressors assessed in environmental impact assessments (EIA) are stipulated in the “environmental impact assessment law” in Japan (see Fig. B.1); • As for nuclear power plants, they are raised as special cases regarding the application of the law;
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• The special provisions are also stipulated in the “electricity utilities industry law” in Japan. All the INSs will meet the criteria set by the regulatory authorities when they are deployed; • The stressors in the EIA are evaluated for the specific site. (Not site-generic evaluation).
(2) Concerning UR1.1, it is doubtful that some of the items (i.e., society, quality of life, biodiversity, psychology, property values, politics and infrastructure) which are not evaluated in Japan’s EIA should be evaluated here, because evaluating such stressors is very difficult and the meaning of the evaluation is not clear. If necessary, the meaning of these items should be mentioned in the present manual. (3) The manual should be prepared so as to assess the potential conformity of the INS to the environment according to the instruction manual. (The present manual refers to a lot of documents and shows their studies as examples in many pages; however, it is difficult for our assessors to understand the manual without reading these references.) (4) The manual should distinguish the items to be assessed according to the maturity level (site selected or not, preliminary, conceptual or final design/ construction/ operation etc.) in the development of INS, because the items to be assessed may be limited in the early design stage of INS. (5) The current name of BP2 should be revised into an appropriate one to understand the content from the name, for example, “Efficient use of non-renewable resources.” (6) It should be cleared that what the meaning of “off-normal events” (mentioned in the present manual on page 9) is. If the “off-normal events” equals to “Abnormal events,” it should be excluded from the environmental assessment because they are already considered as the “Abnormal events” in the safety analysis. (7) Other comments are shown in the Table B.4.
An Environmental Impact Assessment (EIA) for nuclear facilities is carried out with two phases in Japan. Procedure for “Environmental Impact Assessment (EIA) Law” 1. Selection of Planned Construction Site 2. Environmental Survey EIA Procedure 1 3. Implementation of Environmental Impact Assessment (EIA)-- Assessment items 4. Environmental Impact Statement (EIS) ------related to EIA (see page. 8)
Procedure for “Atomic Energy Basic Law,” “Law for the Regulations of Nuclear Sources material, Nuclear Fuel Material and Reactor,” and/or “Electric Utilities Industry Law” EIA Procedure 2 1. Application for Nuclear Facility Establishment ------Assessment items 2. Safety Review related to EIA (see page. 9) 3. Permission of Nuclear Facility Establishment
4. Application for Approval of the Construction Plan 5, 6, 7 - - -
FIG. B.1a. Outline of the EIA law in Japan.
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EIA Procedure 1 in Japan Procedure for “Environmental Impact Assessment (EIA) Law” 1. Maintenance of favorable conditions for natural structural components of the environment - Atmospheric environment [Air quality (Air pollution), Noise, Vibration, Odor, Other (Wind damage, Low-frequency air vibration, etc.)] - Water environment [Water quality, Sediment quality, Underground water, Other (Thermal effluent, Flowing quantity of river, Water circulation, etc.)] -Soiland other environments [Topography, Geography, Ground base, Soil, Other (Sunshiny obstruction, Light pollution, etc.)] 2. Ensuring biodiversity and systematic preservation the natural environment - Flora [Important seed, Habitat that should be paid attention] - Fauna [Important seed, Important colony] - Ecosystem [Ecosystem that characterizes the region] 3. Rewarding contact between people and nature - Scenery [The main view point, Scenery resource, The main view scenery] -Placeof contact [Locations for activities providing contact with nature (Wooden area, waterside and mountain around residence ground)] 4. Environmental load - Water, etc. [Waste, Construction by-product] - Greenhouse gases, etc. [CO , CH ] 2 3 FIG. B.1b. Outline of the EIA law in Japan (procedure 1).
EIA Procedure 2 in Japan Application for Nuclear Facility Establishment Permission
Evaluation items in the attachment document 6 1. Site 2. Weather 3. Ground 4. Underground waterway 5. Earthquake 6. Social environment
Evaluation items in the attachment document 9
- Radiation exposure management - Radiation exposure evaluation at normal operation - Disposal of radioactive waste FIG. B.1c. Outline of the EIA law in Japan (procedure 2).
FIG. B.1d. Outline of the EIA law in Japan (INPRO approach).
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TABLE B.4. RESULTS OF INPRO ENVIRONMENT CRITERIA EVALUATION
BP: Basic Principle BP1: Acceptability of Expected Adverse Environmental Effects UR: User Requirement UR1.1: Controllability of Environmental Stressors CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR1.1.1 : IN1.1.1 : LSt-i, AL1.1.1: LSt-i < Yes level of stressor Si, The radiological stressors evaluated in i. where Si is the the design study of FS met the standard for standards, and other stressors which stressor i. will be evaluated in the EIA shall meet the standards in Japan. [Comments] -The Environmental Impact Assessment (EIA) for nuclear facilities is carried out for two phases in Japan. Almost all of the stressors mentioned here will be evaluated in the application process for an INS establishment. - The relationship between the INPRO methodology and the “Basic Environmental Law” (1993) established by UN should be mentioned. (see pages 6-9) BP: Basic Principle BP1: Acceptability of Expected Adverse Environmental Effects UR: User Requirement UR1.2: Adverse Effects ALARP (As low as reasonably practicable) CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR1.2.1 : IN1.2.1: Does AL1.2.1: Yes Yes the INS reflect In the FaCT Project, ALARA (Similar the application concept to ALARP) will be applied to of ALARP to the R&Ds. limit environmental effects?
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TABLE B.4. RESULTS OF INPRO ENVIRONMENT CRITERIA EVALUATION (CONTINUED)
BP: Basic Principle BP2: Fitness for Purpose UR: User Requirement UR2.1: Consistency with Resource Availability CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR2.1.1 : IN2.1.1: Fj (t) : AL2.1.1: Fj (t) > 0 Yes Consistency quantity of ∀ t < 100 years - FBR cycle in the FaCT Project will with fissile/fertile be able to contribute to the world’s Fissile/Fertile material j energy needs during the 21st century Resource available for use without running out of fissile/fertile Availability in the INS at material. time t. [Comments] - It is meaningless for each country to evaluate Fj(t) because it depends largely on many factors such as INS installation rate and capability of conventional nuclear facilities. CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR2.1.2 : IN2.1.2: Qj (t) : AL2.1.2: Qj (t) > Yes Consistency quantity of 0 - Non-renewable resources are not with Material material i ∀ t < 100 year evaluated in the Japan’s FaCT design Resource available for use study. However, key, critical or Availability in the INS at strategic non-renewable resources are time t. not needed for the FaCT project. [Comments] - Non-renewable resources to be evaluated should be limited to key, critical or strategic non-renewable resources. CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR2.1.3 : IN2.1.3: P (t): AL2.1.3: P(t) Not evaluated in the design study of Consistency power available ≥PINS(t) the FaCT. with Power (from both ∀t < 100 years, [Comments] Availability internal and where PINS(t) is - Every INS must meet this external the power requirement because we are sources) for use required by the developing energy generation systems, in the INS at INS at time t. therefore this AL is nonsense. time t. CR: Criteria Evaluation results
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CR: Criterion IN: Indicator AL: Acceptance Limit CR2.1.4 : IN2.1.4: U : end AL2.1.4: U > U0 Yes Consistency use (net) energy U0 : maximum - U0 is estimated to be about 600PWh with Resource delivered by the achievable for a (6x1017Wh) under the following Availability: INS per Mg of once through assumption: available natural U Net Energy U uranium mined. PWR. resources: 14,800,000 tU and once- through LWR and HWR. [Comments] - U availability (Utilization factor of U) as an indicator is better than Net Energy U. CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR2.1.5 : IN2.1.5: T : end AL2.1.5: T > T0 Not evaluated in the design study of Consistency use (net) energy T0 : maximum T the FaCT. with Resource delivered by the achievable with a Availability: INS per Mg of current operating Net Energy Th thorium mined. thorium cycle. CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR2.1.6 : IN2.1.6: Ci : end AL2.1.6: Ci > C0 Not evaluated in the design study of Consistency use (net) energy C0 to be the FaCT. with Resource delivered per determined on a [Comments] Availability: Mg of limited case specific The purpose is unknown. Net Energy non-renewable basis. nR, resource nR: non- consumed renewable BP: Basic Principle BP2: Fitness for Purpose UR: User Requirement UR2.2: Adequate Net Energy Output CR: Criteria CR: Criterion IN: Indicator AL: Acceptance Evaluation results Limit CR2.2.1 : IN2.2.1: T EQ : AL2.2.1: T EQ < k Not evaluated in the design study of time required to ·TL the FaCT. match the total TL : intended life [Comments] energy input of INS k < 1 - Energy for decommissioning should with energy be included into the energy input. output (years). - The Energy Profit Ratio (EPR) is rather suitable for UR 2.2 because TEQ largely depends on an introducing rate of INSs, etc. and it is very difficult to evaluate it.
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- O. Amano, CREIPI estimates the EPRs for LWR and FBR cycles: 17.6 for LWR cycle, around 30-40 for FBR cycle. (J. of Atomic Energy Society of Japan, Vol.48, No.10, 2006, in Japanese)
B.4. SAFETY An outline of the JSFR safety features is presented in this section. B4.1. Inherently safe characteristics, robustness, and inertia to cope with transients Unlike light water reactors, the coolant pressure is kept at ~1atm and the boiling point of sodium at this pressure is very high compared with the operating temperature, therefore providing a high inertia to cope with temperature-increase transients. There is also no pressure-increase transient in the JSFR as there is no rapid flashing in the primary coolant system. Further, there is no potential for a loss-of coolant accident (LOCA) within design basis accidents (DBA) (Fig. B.2) and it is easy to maintain coolant inventory even after a boundary failure. In addition, there is no reactivity insertion accident due to abrupt ejection of a control rod owing to the low pressure system. Design simplicity in the accident prevention systems In order to obtain robustness and reduce incident and/or accident frequency of the JFSR, design simplicity is pursued (Fig. B.2): • Elimination of piping penetrations at the bottom and lateral side of both the reactor vessel and the guard vessel; • Reducing the loop number in the main cooling system to two; • Reducing small-size piping and shortening total piping length in the main cooling system combined with enhancing integrity at a high temperature by adopting high-chromium steel for the coolant boundary; • Integration of the primary pump and intermediate heat exchanger; • Inclusion of the auxiliary sodium system to the inside of the main cooling system. Specific issues to sodium There are two specific issues related to sodium (Figs B.2and B.3). One issue is flammability. We are pursuing a localization of the damage due to sodium chemical reaction by introducing passive safety features such as guard pipes and guard vessels that are filled with inert gas in the primary sodium system, and high reliable steam generators that are equipped with double-wall heat transfer tubes in order to prevent intensive Na-H2O chemical reaction. The other issue is accommodation of in service inspection and repair requirement. We are planning to clarify the inspection methods and technologies to be developed for sodium cooling systems (e.g., remote handling sensor under sodium at 200℃).
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Passive safety features for reactor shutdown, makeup of coolant inventory and decay heat removal Reactor core safety in the JSFR is enhanced by passive safety features (Fig. B.2and B.4). In order to prevent a core disruptive accident (CDA) even upon failure of the active reactor shutdown devices (i.e., detectors, reactor trip signal transmission circuits, reactor trip breakers), the Curie point electromagnet type self-actuated shutdown system (SASS) is introduced in the backup reactor shutdown system. A key feature of the SASS is a passive de-latch mechanism utilizing magnetic resistance change at the Curie-point temperature of the sensing alloy (30Ni-32Co-Fe). The R&D of the SASS utilizing the JOYO is in progress to enable the prevention of a control rod drop due to inadvertent actuation of the SASS with high reliability. There has been no sign of spurious drops so far. Even if the SASS fails, the sodium void reactivity is limited to less than 6 $, thus eliminating a mechanistic energy release due to a core excursion in the initiating phase of a CDA progression. In addition, after the initiating phase, mitigation of a CDA is achieved by introducing a fuel assembly with inner duct structure (FAIDUS) ‒ that is a fuel assembly design for enhancement of molten fuel discharge after an unprotected core degradation ‒ and by installing an in-vessel core catcher. The DHRS is capable of initiating natural circulation (i.e. no pump and valve needed) under loss of off-site power. In addition, decay heat can be removed actively by means of the primary and secondary heat transport system combined with the main feed-water and main steam system if off-site power is available. B4.2. Evaluation results According to the IAEA-TECDOC-1434, titled “Methodology for the assessment of innovative nuclear reactors and fuel cycles”, in the area of safety for innovative reactors and fuel cycle facilities a set of basic principles, user requirements, and criteria (i.e., indicators and acceptance limits) has been defined. There are 4 safety basic principles (BP1 to BP4) in total. As for BP1, 7 user requirements (UR) and 22 indicators are defined; a single user requirement and 4 indicators are defined for BP2; 2 user requirements and 2 indicators for BP3; and 4 user requirements and 10 indicators for BP4. Acceptance limits are also defined corresponding to each indicator. The definitions of all the BP, UR, indicators and acceptance limits are shown in Table B.5 together with evaluation results of the JSFR. In the present section, typical evaluation results corresponding to UR1.1 through UR1.5 and UR2.1 are described. The other items have no specific issues related to the JSFR and can be achieved by adopting a standard design approach same as the existing nuclear systems (see Table B.5). INPRO user requirement UR1.1 robustness Concerning IN1.1.1 “Robustness of design”, the JSFR is superior to existing designs in (1) the design simplicity in the prevention systems, (2) robustness, and (3) improved materials as described in Section 4.1.1 and 4.1.2. Concerning IN1.1.2 “High quality of operation”, the JSFR is superior to existing designs in (1) appropriate operational margins to reduce unnecessary scrams, (2) reducing scram frequency owing to design simplicity (see Section 4.1.2), and (3) use of operating experience of other NPPs. In particular, it should be noted that we have been making use of the operating experience
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obtained from “JOYO”, “MONJU”, and R&D of the sodium-related components and systems as well as the existing LWRs and that the experienced MONJU personnel have always participated in the design work related to plant operation and maintenance. Concerning IN1.1.3 “Capability to inspect” and IN1.1.4 “Expected frequency of failures and disturbances”, the JSFR is superior to existing designs as shown in Section 4.1.3 and Sections 4.1.2 and 4.1.4, respectively. Concerning IN1.1.6 “Inertia to cope with transients”, the JSFR is superior to existing designs as shown in Section 4.1.1.
INPRO user requirement UR1.2 detection and interception of deviations from normal conditions The JSFR satisfies the acceptance limits of IN1.2.1 “Capability of I&C system / inherent characteristics”. This is because in case of abnormal deviations the reactor shuts down and decay heat is removed safely and automatically by active combined with passive safety features before progressing to an accident: • A series of the passive shutdown actuation mechanism of the SASS installed in the backup RSS; • DHRS is capable of natural circulation (no pump and valve needed).
INPRO user requirement UR1.3 reduction of accident frequency, restoration to a controlled state, minimal reliance on human intervention The JSFR satisfies the acceptance limit of IN1.3.1 “Calculated frequency of occurrence of design basis accidents” as described in Sections 4.1.1 through 4.1.3. The acceptance limits of IN1.3.2 “Grace period until human intervention (> 8 hours)” are satisfied. This is because the automatic safety systems operate on electric power supply from batteries and emergency generators even under the DBA + LOSP conditions so that no operator action is required during 3.5 days. Concerning IN1.3.3 “Reliability of engineered safety features”, the high reliability of the engineered safety features has been verified by performing a preliminary level-1 PSA. In other words, a very low likelihood of failure of all confinement barriers has been demonstrated by a low core damage frequency in the preliminary level-1 PSA (See IN1.4.1) The number of remaining barriers is deterministically designed to be 1 at least (in case of primary coolant leakage). Thus, the acceptance limit of IN1.3.4 “Number of confinement barriers maintained” is satisfied. In addition, as described in Sections 4.1.4 and 4.2.2, the acceptance limit of IN1.3.5 “Capability of the engineered safety features to restore the INS to a controlled state (without operator actions)” is also satisfied. INPRO user requirement UR1.4 reduced major release frequency of radionuclides in the confinement (due to internal events) and UR1.5 major release of radioactivity prevented, no evacuation Only a severe core damage event can lead to a major release of radionuclides into the confinement. The calculated core damage frequency has been ~3.4x10-8/a (ATWS: 1x10-8/a, LORL: 4x10-9/a, PLOHS: 2x10-8/a) based on the preliminary level 1 PSA. As described in Section 4.1.4, there are measures for controlling severe accidents in confinement including in- plant severe accident management measures such as Curie point electromagnet type self-actuated shutdown system (Fig. B.3), fuel assemblies with inner duct structure (Fig. B.4), and in-vessel
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core catcher for the unprotected sequences (=ATWS); manual operation of the air cooler dampers in long-term LOSP, additional installation of the backup air cooler dampers with diversity, and auxiliary steam supply from water tank into steam generators. Thus, the acceptance limits of the indicators IN1.4.1 through IN1.4.3 are satisfied. The major release frequency of radionuclides into the environment could become less than 3.4x10-8/a of the core damage frequency. INPRO basic principle BP2/UR2.1 Inherently safe characteristics and passive systems As described in Section 4.1.1, 4.1.3 and 4.1.4, there are many inherently safe characteristics and passive systems in the JSFR so that the JSFR can satisfy the acceptance limits of IN2.1.1.
(1) Stored energy: Unlike LWRs, the coolant pressure is ~1atm; there is no rapid flashing and it is easy to maintain coolant inventory upon the boundary failure; (2) Core coolability: Natural circulation of the Na coolant; (3) Reactivity feedback: a. Negative power coefficient within DBA; b. Passive reactor shutdown by SASS against beyond-DBA; c. Even if SASS fails, the limited sodium void reactivity < 6 $ so as to eliminate mechanistic energy release due to core excursion in the initiating phase of the CDA progression; d. Re-criticality-free core: FAIDUS. (4) Flammability of Na: Pursuing localization of the damage with passive safety features. (e.g., guard pipe, double-wall-tube SG, etc.). Summary of the INPRO assessment in the area of safety The design concept of the JSFR is simple and robust, and has enhanced passive safety features. In the safety design and evaluation, deterministic and probabilistic approaches are employed. We are planning and implementing RD&D activities for the innovative technologies. For all the user requirements and indicators, the acceptance limits are or would be satisfied. B4.3. Comments to INPRO methodology • Comments on reactor safety assessment manual (1) It is recommended to use technical terms in a consistent manner and clarify the definition of the technical terms: o IN2.1.2 (expected frequency of abnormal operation and accidents) in comparison with IN1.1.4 (expected frequency of failures and disturbances) and IN1.3.1 (calculated frequencies of occurrence of design basis accidents); o IN2.1.3 (consequences of abnormal operation and accidents) in comparison with IN1.5.2 (calculated consequences of release), IN3.1.1 (occupational dose values) and IN3.2.1 (public dose values); o The inventory of “radioactive materials” mentioned in IN2.1.1. (2) Concerning the simplicity and margins discussed in IN1.1.1 (i.e., robustness of design), if there is an excessive margin in some part of the existing reactor, it should be also regarded as “superior” to reallocate the excessive margin so as to optimize both the entire safety and the other viewpoints; e.g., shortening the piping brings about both the simplicity in prevention
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system (i.e., reduction of system failure rate) and the capital cost reduction while the structural reliability margin might be relatively reduced. (3) Concerning IN1.1.2 (i.e., high quality of operation) and IN1.1.4 (i.e., expected frequency of failures and disturbances), it is better to show us the detailed guideline of the estimation methodology. Otherwise, practically it might be better to require the evaluation of this indicator after INS enters the operating phase. This is because no one can prove the reliability of the innovative components/devices equal to or superior to that of the existing ones on the basis of the classic statistical calculation. • Comments on safety assessment manual for reprocessing facility (1) It is recommended to use technical terms in a consistent manner and clarify the definition of the technical terms. “Major release” in IN1.4.1 (calculated frequency of major release of radioactive materials into the containment/confinement) and IN1.5.1 (calculated frequency of major release of radioactive materials to the environment) in terms of the difference from IN1.3.1 (calculated frequencies of occurrence of design basis accidents). (2) Concerning the items shown below, it is better to add more explanations in order to obtain better understanding of: o The basis of the quantitative acceptance limits of IN1.4.1 (<10-4/a); o The basis of the quantitative acceptance limits of IN1.5.2 (<1 mSv); o The air operated motor (appears in the explanation of UR2.1).
BP1 / UR1.1 & BP2 / UR2.1 No LOCA potential owing to high boiling point at ~1atm
Pursuit of design simplicity for robustness & reduction BP2 / UR2.1 of incident & accident BP1 / UR1.1 / IN1.1.3 frequency
BP1 / UR1.1, UR1.3 BP1 / UR1.2～UR1.5 & BP2 / UR2.1 FIG. B.2. Main safety features of JSFR.
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BP2 / UR2.1 inherent & passive safety
BP1 / UR1.1 / IN1.1.3 Inspection
FIG. B.3. Resolution of specific issues related to sodium.
FIG. B.4. Enhanced safety features of the JSFR.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION
Safety Basic Installations of an Innovative Nuclear Energy System shall incorporate Principle BP1 enhanced defence-in-depth as a part of their fundamental safety approach and ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installations. User Installations of an INS should be more robust relative to existing designs Requirement: regarding system and component failures as well as operation. UR1.1 Criteria Evaluation results IN: Indicator Acceptance Limit IN1.1.1: Superior to Yes. Robustness of existing (1) The design simplicity of the prevention systems design designs in at least • No piping penetration at the RV lateral side. (simplicity, some of the • Reducing the loop number in the main cooling margins). aspects discussed system to 2. in the text. • Shortening the total piping length in the main cooling system. (2) Robustness • No LOCA potential within DBA because of high boiling point of “Na” coolant nearly at the atmospheric pressure. (3) Improved materials • Enhancing integrity at a high temperature (a) High-chromium steel for the coolant boundary (b) Oxide dispersion strengthened steel for the fuel cladding IN1.1.2: High Superior to Yes. quality of existing (1) Appropriate operational margins to reduce operation. designs in at least unnecessary SCRAM: Set the adequate trip level in some of the any plant parameters such as neutron flux, coolant aspects discussed temperature, etc. in the text. (2) Reduce scram frequency owing to design simplicity (See IN1.1.1 (1)) (3) Use of Operating Experience of Other NPPs Obtained from “JOYO”, “MONJU”, and R&D of the sodium-related components and systems as well as the existing LWRs. The experienced MONJU personnel have participated in the design work related to the plant operating and maintenance.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
Safety Basic Installations of an Innovative Nuclear Energy System shall incorporate Principle BP1 enhanced defence-in-depth as a part of their fundamental safety approach and ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installations. User Installations of an INS should be more robust relative to existing designs Requirement: regarding system and component failures as well as operation. UR1.1 Criteria Evaluation results IN: Indicator Acceptance Limit IN1.1.3: Superior to Yes. Capability to existing (1) Plan to clarify the inspection methods and inspect. designs in at least technologies to be developed for the sodium cooling some of the systems. aspects discussed (2) Plan to establish the framework of inspection and in the text. maintenance strategies for the JSFR considering the latest technologies. IN1.1.4: Superior to Yes. (See IN1.1.1 (1), IN1.1.2 (2)) Expected existing (1) The pump trip frequency in the main cooling frequency of designs in at least systems lower than that of 3-loop or 4-loop cooling failures and some of the systems. disturbances. aspects discussed (2) The sodium leakage event frequency is minimized in the text. by shortening the piping and simplifying the auxiliary sodium system (i.e., inclusion to the inside of the main cooling system). (3) Control rod drop due to incorrect actuation of the SASS: The R&D of the SASS utilizing the JOYO is in progress so that this can be prevented with high reliability. No sign of spurious drops, so far. IN1.1.5: Grace Superior to Yes. period until existing (1) Not necessary at all interventions by the operator human actions designs in at least against abnormal transients because the RSS and the are required. (> some of the DHRS automatically function. 30 min) aspects discussed (2) The operator can perform appropriate actions in the text. practically without time limits and under less stressed conditions when a slight and slow transient occurs. IN1.1.6: Inertia to Superior to Yes. cope with existing (1) Unlike LWRs, there is no pressure-increase transients. designs in at least transient in the JSFR so that no flow comes out of the some of the primary coolant system. Therefore, there never results aspects discussed in the situation of delaying a return to normal in the text. operation. (2) High inertia to cope with the temperature-increase transients because of high boiling point of sodium coolant. （ See IN1.1.1(2))
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
Safety Basic Installations of an Innovative Nuclear Energy System shall incorporate Principle BP1 enhanced defence-in-depth as a part of their fundamental safety approach and ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installations. User Installations of an INS should detect and intercept deviations from normal Requirement: operational states in order to prevent anticipated operational occurrences UR1.2 from escalating to accident conditions Criteria Evaluation results IN: Indicator Acceptance Limit IN1.2.1: Key system Yes. Capability of variables relevant Upon abnormal deviations, the reactor shuts down and control and to safety (e.g. decay heat is removed safely automatically before instrumentation flow, pressure, progressing to an accident. system and/or temperature, A series of the passive shutdown actuation inherent radiation levels) mechanisms of the SASS installed in the backup characteristics to do not exceed RSS. detect and limits acceptable DHRS is capable of natural circulation (no pump intercept and/or for continued and valve needed). compensate such operation (no deviations. event reporting necessary). User The frequency of occurrence of accidents should be reduced, consistent Requirement: with the overall safety objectives. If an accident occurs, engineered safety UR1.3 features should be able to restore an installation of an INS to a controlled state, and subsequently (where relevant) to a safe shutdown state, and ensure the confinement of radioactive material. Reliance on human intervention should be minimal, and should only be required after some grace period. Criteria Evaluation results IN: Indicator Acceptance Limit IN1.3.1: Reduced Yes. Calculated frequency of (1) The pump sticking accident reduced by reducing frequency of accidents that can the number of pumps. occurrence of cause plant (2) The SGTR accident reduced by adopting a double- design basis damage relative to wall tube type SG. (Prevent intensive Na-H2O accidents. existing facilities. chemical reaction) (3) The primary coolant leakage accident reduced by establishing the LBB logic, shortening the main piping, and reducing the small-size piping. (4) Unlike LWRs, there is no reactivity insertion accident due to abrupt ejection of the control rod.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
Safety Basic Installations of an Innovative Nuclear Energy System shall incorporate Principle BP1 enhanced defence-in-depth as a part of their fundamental safety approach and ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installations. User The frequency of occurrence of accidents should be reduced, consistent Requirement: with the overall safety objectives. If an accident occurs, engineered safety UR1.3 features should be able to restore an installation of an INS to a controlled state, and subsequently (where relevant) to a safe shutdown state, and ensure the confinement of radioactive material. Reliance on human intervention should be minimal, and should only be required after some grace period. Criteria Evaluation results IN: Indicator Acceptance Limit IN1.3.2: Grace Increased relative Yes. period until to existing Even under the DBA + LOSP condition, no operator human facilities. (> 8 action necessary during 3.5 days because the intervention is hours) automatic safety systems function by the electric necessary. power supply from the batteries and the emergency generators. IN1.3.3: Equal or superior Yes. Reliability of to existing Verified the high reliability of the engineered safety engineered safety designs. features by performing the preliminary level-1 PSA. features. IN1.3.4: Number At least one. Yes. of confinement The number is deterministically designed to be at least barriers 1 (in case of primary coolant leakage). maintained. A very low likelihood of failure of all barriers is caused by low frequency of core damage in the preliminary level-1 PSA (See IN1.4.1) IN1.3.5: Sufficient to reach Yes. Capability of the a controlled state. (1) Reactor scram and decay heat removal: same engineered safety response as IN1.2.1. features to restore (2) In case of the coolant leakage, the required reactor the INS to a coolant level is made up by the safety functions: guard controlled state vessel and guard pipe, reactor cover gas isolation. (without operator Leaked sodium is confined inside the guard vessel / actions). guard pipe. No challenge to the containment.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
IN1.3.6: Sub- Sufficient to cover Yes. criticality uncertainties and The shutdown reactivity margin of the primary RSS is margins allow adequate 1.7 %Δk/k’ according to the core design calculation grace period. (> considering the uncertainties. 1 %Δk/k’). The primary RSS alone can attain cold shutdown from the full-power operation even in the case of sticking a single control rod with the highest worth. User The frequency of a major release of radioactivity into the containment / Requirement: confinement of an INS due to internal events should be reduced. Should a UR1.4 release occur, the consequences should be mitigated. Criteria Evaluation results IN: Indicator Acceptance Limit IN1.4.1: At least an order Yes. Calculated of magnitude less The calculated core damage frequency is ~3.4x10-8/a frequency of than for existing (ATWS: 1x10-8/a, LORL: 4x10-9/a, PLOHS: 2x10-8/a) major release of designs; even based on the preliminary level 1 PSA. radioactive lower for materials into the installations at containment / urban sites. confinement. IN1.4.2: Natural Existence of such Yes. or engineered processes. (1) For the unprotected sequences (=ATWS): Curie processes point electromagnet type self-actuated shutdown sufficient for system (SASS), Fuel Assemblies with Inner DUct controlling Structure (FAIDUS), in-vessel core catcher. relevant system (2) For the protected sequences (=LORL+PLOHS): parameters and See IN1.4.3. activity levels in containment / confinement IN1.4.3: In-plant Procedures, Yes. Examples of AM for prevention of severe severe accident equipment and accident management training sufficient • Manual operation of the air cooler (AC) dampers in to prevent large long-term LOSP release outside • Additional installation of the backup AC dampers containment / with diversity confinement and • Auxiliary steam supply from water tank into SGs regain control of the facility.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
User A major release of radioactivity from an installation of an INS should be Requirement: prevented for all practical purposes so that INS installations would not need UR1.5 relocation or evacuation measures outside the plant site, apart from those generic emergency measures developed for any industrial facility for similar purposes. Criteria Evaluation results IN: Indicator Acceptance Limit IN1.5.1: Calculated Yes. Calculated frequency <10-6 It could become < 3.4x10-8/a of CDF (See IN1.4.2 & frequency of a per unit-year, or IN1.4.3) so that practically there is no major major release of practically containment failure. radioactive excluded by materials to the design. environment. IN1.5.2: Consequences Yes. Calculated sufficiently low to Practically no major containment failure. consequences of avoid necessity for releases (e.g. evacuation. dose). Appropriate off- site mitigation measures (e.g. temporary food restrictions) are available. IN1.5.3: Comparable to Already shown that nuclear energy production has a Calculated facilities used for low risk based on the “Years of Life Lost (YOLL)”, individual and similar purposes. which is referred to in the INPRO user’s manual. collective risk. User An assessment should be performed for an INS to demonstrate that the Requirement: different levels of defence-in-depth are met and are more independent from UR1.6 each other than for existing systems. Criteria Evaluation results IN: Indicator Acceptance Limit IN1.6.1: Adequate Yes. Independence of independence is Established a framework of safety assurance for the different levels of demonstrated, e.g. JSFR constituting the structure of DID, combined with DID through three fundamental safety functions (i.e. reactivity deterministic and control, core cooling and containment of probabilistic radionuclides). means, hazards analysis etc.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
User Safe operation of installations of an INS should be supported by an Requirement: improved Human Machine Interface resulting from systematic application UR1.7 of human factors, requirements to the design, construction, operation and decommissioning. Criteria Evaluation results IN: Indicator Acceptance Limit IN1.7.1 : Satisfactory results Yes. Evidence that from assessment. The JSFR will adopt the state-of-the-art methodology human factors regarding human factors (HF). (HF) are addressed systematically in the plant life cycle. IN1.7.2 : - Reduced Yes. Application of likelihood of The HF models and artificial intelligence technologies formal human human error to be used for the JSFR will be state-of-the-art which response models relative to existing implies data from other industries that are used to the from other plants, as extent possible. industries or predicted by HF development of models. nuclear specific - Use of artificial models intelligence for early diagnosis and real-time operator aids - Less dependence on operator for normal operation and short term accident management relative to existing plants
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
Safety Basic Installations of an INS shall excel in safety and reliability by incorporating Principle BP2 into their designs, when appropriate, increased emphasis on inherently safe characteristics and passive systems as a part of their fundamental safety approach. User INS should strive for elimination or minimization of some hazards relative Requirement: to existing plants by incorporating inherently safe characteristics and/or UR2.1 passive systems, when appropriate. Criteria Evaluation results IN: Indicator Acceptance Limit IN2.1.1: Sample Superior to Yes. indicators: stored existing designs. (1) Stored energy: Unlike LWRs, the coolant pressure energy, is ~1atm; there is no rapid flashing and it is easy to flammability, maintain coolant inventory upon the boundary failure. criticality, (See IN1.1.1 (2), IN1.3.5 (2)) inventory of (2) Core coolability: Natural circulation of the Na radioactive coolant materials, (3) Reactivity feedback: available excess a. Negative power coefficient within DBA. reactivity, b. Passive reactor shutdown by SASS against reactivity beyond-DBA. feedback. c. Even if SASS fails, the sodium void reactivity limited < 6$ so as to eliminate mechanistic energy release due to core excursion in the initiating phase of the CDA progression. d. Re-criticality-free core: FAIDUS (4) Flammability of Na: Pursuing localization of the damage with passive safety features. (e.g., guard pipe, double-wall-tube SG, etc.) IN2.1.2: Lower frequencies Yes. Expected compared to See IN1.1.4 & IN1.3.1. frequency of existing facilities. abnormal operation and accidents. IN2.1.3: Lower consequences Yes. Consequences of compared to existing See IN1.2.1 & IN1.3.5. abnormal operation facilities. and accidents. IN2.1.4: Confidence Validity established. Yes. in innovative The R&D works for ensuring the confidence: completed/on- components and going. The R&D plans for establishing the technology base: approaches. developed / launched. The demonstration plan of the JSFR technology is being developed.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
Safety Basic Installations of an INS shall ensure that the risk from radiation exposures to Principle BP3 workers, the public and the environment during construction/commissioning, operation, and decommissioning, are comparable to the risk from other industrial facilities used for similar purposes. User INS installations should ensure an efficient implementation of the concept Requirement: of optimization of radiation protection through the use of automation, UR3.1 remote maintenance and operational experience from existing designs. Criteria Evaluation results IN: Indicator Acceptance Limit IN3.1.1: Less than the Yes. Occupational limits defined by Preliminary estimation of representative components dose values. national laws or should be done. international Some appropriate measures (e.g., shielding) should be standards so that taken in the detailed design work. the health hazard to workers is comparable to that from an industry used for a similar purpose. User Dose to an individual member of the public from an individual INS Requirement: installation during normal operation should reflect an efficient UR3.2 implementation of the concept of optimization, and for increased flexibility in siting may be reduced below levels from existing facilities. Criteria Evaluation results IN: Indicator Acceptance Limit IN3.2.1: Public Less than the Yes. dose values. limits defined by Preliminary result of ~1μSv/a/GW(e), by assuming national laws or 1% fuel pin defect. international standards so that the health hazard to the public is comparable to that from an industry used for a similar purpose
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
Safety Basic Installations of an INS shall excel in safety and reliability by incorporating Principle BP4 into their designs, when appropriate, increased emphasis on inherently safe characteristics and passive systems as a part of their fundamental safety approach. User The safety basis of INS installations should be confidently established prior Requirement: to commercial deployment. UR4.1 Criteria Evaluation results IN: Indicator Acceptance Limit IN4.1.1: Safety Yes for all. Yes. Prepared well-documented draft version: concept defined. (1) safety design requirements IN4.1.2: Design- (2) safety evaluation rules related safety (3) major safety evaluation results requirements specified. IN4.1.3: Clear process for addressing safety issues. User Research, Development and Demonstration on the reliability of components Requirement: and systems, including passive systems and inherent safety characteristics, UR4.2 should be performed to achieve a thorough understanding of all relevant physical and engineering phenomena required to support the safety assessment. Criteria Evaluation results IN: Indicator Acceptance Limit IN4.2.1: RD&D Yes Yes. Major RD&D already identified; discussing the defined and individual conduction plan. Reliability database performed, and CORDS for SFR’s PSA. database developed. IN4.2.2: Yes Yes. Already done & applied in the various safety- Computer codes related regions. or analytical methods developed and validated. IN4.2.3: Scaling Yes Yes. Discussing scaling effect. understood and/or full scale tests performed.
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TABLE B.5. INPRO CRITERIA AND RESULTS OF SAFETY EVALUATION (CONTINUED)
User A reduced-scale pilot plant or large-scale demonstration facility should be Requirement: built for reactors and/or fuel cycle processes, which represents a major UR4.3 departure from existing operating experience. Criteria Evaluation results IN: Indicator Acceptance Limit IN4.3.1: Degree -High degree of High degree of novelty. For all innovative features, of novelty of the novelty: Facility RD&D are underway. process. specified, built, operated, and lessons learned, documented. -Low degree of novelty: Rationale provided for bypassing pilot plant. IN4.3.2: Level of Results sufficient Discussing demonstration reactor plan. adequacy of the to be extrapolated. pilot facility. User For the safety analysis, both deterministic and probabilistic methods should Requirement: be used, where feasible, to ensure that a thorough and sufficient safety UR4.4 assessment is made. As the technology matures, “Best Estimate (plus Uncertainty Analysis)” approaches are useful to determine the real hazard, especially for limiting severe accidents. Criteria Evaluation results IN: Indicator Acceptance Limit IN4.4.1: Use of a Yes Yes. Preliminary PSA applied (power operation; risk informed simplified manner). approach. IN4.4.2: Yes Yes. Sensitivity studies will correspond to the degree Uncertainties and of detail in the design work. sensitivities identified and appropriately dealt with.
B5. WASTE MANAGEMENT
B5.1. Introduction Japan’s studies on the radioactive waste management area in the INPRO Joint Study were carried out based on the INPRO manual. Japan has started the FaCT project to research and develop, and demonstrate commercialized FR (Fast Reactor) cycle systems for the future. Since the FR cycle
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which consists of the JSFR and its fuel cycle facilities was selected as the principal concept of the project, it was also selected as the reference system in this study. Quantitative data for the reference system should be used in the evaluation if they are available. Otherwise data for existing LWR (Light Water Reactor) cycle systems or engineering judgments based on the characteristics of the system and experiences in the waste management area could be used in the evaluation. Consequently, quantitative figures and descriptions were mixed in the evaluation result. B5.2. Evaluation results The radioactive waste management area of INPRO has four basic principles (BP1 to BP4) and seven user requirements (URs). Each BP has two user requirements (except BP1 has only one). The BPs and URs are as follows: • BP1 Waste minimization: o UR1.1 reduction of waste at the source. • BP2 Protection of human health and the environment, o UR2.1 Protection of Health; o UR2.2 Protection of Environment. • BP3 Burden on future generations: o UR3.1 End State; o UR3.2 Attribution of Waste Management Costs. • BP4 Waste optimization: o UR4.1 Waste Classification; o UR4.2 Pre-disposal Waste Management. Table B.6 shows the detailed evaluation results of radioactive waste management area in Japan’s studies. On the whole, no problems were found according to the evaluation result in this area for the reference system. The evaluation was conducted based on experiences and knowledge of radioactive waste from existing LWR cycle systems rather than from a future INS which has not been deployed yet. As for the evaluation results of UR1, the quantities and qualities of radioactive wastes from the INS were considered superior than those of the wastes from the existing LWR cycle. Regarding BP2 (UR2.1 and UR2.2), it was concluded that regulatory standards on radioactive wastes will be met when the INS is deployed. Concerning UR3.1 and UR3.2, the technology, resource, time, etc., necessary for the INS were considered as available and the cost for waste management was included in the economics evaluation. As for BP4 (UR4.1 and UR4.2), the same classification scheme for radioactive waste from existing LWR cycle will also be applied to those from INS. The process of intermediate steps to the final end state of radioactive waste are described and technical indicators are expected to be established by the regulatory bodies although there is a comment regarding AL4.2.1 (see the comments in in the following section).
B5.3. Comments to the INPRO methodology
• The way of evaluation will be altered according to progress of the R&D activities. Therefore, they will help users evaluate the waste management area if various sets of criteria
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corresponding to the different R&D stages are presented; • If the EPs (Evaluation Parameters) for the waste management area are presented, it will be beneficial for the users; • ALs for UR1.1 are all “ALARP”, but they will make it difficult for users to evaluate an INS. Typical values for existing LWR cycles may be appropriate as alternative ALs; • AL4.2.1 should be modified because it is not necessary to observe the principle of “as short as reasonably practicable” at the time of producing the waste form specified for the end state.
TABLE B.6. RESULTS OF EVALUATION OF RADIOACTIVE WASTE MANAGEMENT USING THE INPRO METHODOLOGY
BP: Basic Principle BP1: (Waste minimization) Generation of radioactive waste in an INS shall be kept to the minimum practicable. UR: User Requirement UR1.1: (Reduction of waste at the source): The INS should be designed to minimize the generation of waste at all stages, with emphasis on waste containing long- lived toxic components that would be mobile in a repository environment. CR: Criteria IN: Indicator AL: Acceptance Evaluation results and Explanations Limit Alpha-emitters and other ALARP Yes long-lived radio-nuclides per ~3×1013 Bq/GWa （ Alpha-emitters in GWa HLW as manufactured） Total activity per GWa. ALARP Yes (An example for HLW from existing LWR cycle: ~4×1015 Bq/GWa) Mass per GWa. ALARP Yes ~5×1017 Bq/GWa （ Total activity in HLW as manufactured） Volume per GWa. ALARP Yes (An example for HLW from existing LWR cycle: ~6×1017 Bq/GWa) Chemically toxic elements ALARP Yes that would become part of the ~7 t/GWa （ The weight of the glass in radioactive waste per GWa. HLW）
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TABLE B.6. RESULTS OF EVALUATION OF RADIOACTIVE WASTE MANAGEMENT USING THE INPRO METHODOLOGY (CONTINUED)
BP: Basic Principle BP2: (Protection of human health and the environment) Radioactive waste in an INS shall be managed in such a way as to secure an acceptable level of protection for human health and the environment, regardless of the time or place at which impacts may occur. UR: User Requirement UR2.1: (Protection of Human Health) Exposure of humans to radiation and chemicals from INS waste management systems should be below currently accepted levels and protection of human health from exposure to radiation and chemically toxic substances should be optimized. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations (IN2.1.1) Meets regulatory Yes Estimated dose rate to an standards of specific The regulatory standards on the dose individual of the critical group Member State rate will be met when the INSs are deployed. (IN2.1.2) Meets regulatory Yes Radiological exposure of standards of specific The regulatory standards on the workers Member State radiological exposure of workers will be met when the INSs are deployed. (IN2.1.3) Meets regulatory Yes Estimated concentrations of standards of specific The regulatory standards on the chemical toxins in working Member State chemical toxins in working area will areas be met when the INSs are deployed. UR: User Requirement UR2.2: (Protection of the Environment) The cumulative releases of radio-nuclides and chemical toxins from waste management components of the INS should be optimized. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations (IN2.2.1) Meet regulatory Yes Estimated releases of standards of specific Estimated release of the radionuclides and chemical Member State radionuclides and chemical toxins toxins from waste from waste management facilities of management facilities INS will meet the regulatory standards when those facilities are deployed although there is no quantitative data on chemical toxicity at present.
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TABLE B.6. RESULTS OF EVALUATION OF RADIOACTIVE WASTE MANAGEMENT USING THE INPRO METHODOLOGY (CONTINUED)
BP: Basic Principle BP3: (Burden on future generations) Radioactive waste in an INS shall be managed in such a way that it will not impose undue burdens on future generations. UR: User Requirement UR3.1: (End State): An achievable end state should be specified for each class of waste, which provides permanent safety without further modification. The planned energy system should be such that the waste is brought to this end state as soon as reasonably practicable. The end state should be such that any release of hazardous materials to the environment will be below that which is acceptable today. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations (IN3.1.1) All required technology Yes Availability of technology is currently available or The disposal concepts and reasonably expected to technologies described in the H12* be available on a are applicable as a technical basis to schedule compatible the HLWs from the INS in Japan. with the schedule for Furthermore, the existing and introducing the studying technologies for the LLWs proposed innovative from LWR cycle are applicable to fuel cycle those from FR cycle. The specifications of the LLWs may be changed from the differences of the inventories.
H12*: “The Second Progress Report on R&D for the Geological Disposal of HLW in Japan” (IN3.1.2) Any time required to Yes Time Required bring the technology to The technologies developed in H12 the industrial scale must are utilized as a technical basis for be less than the time the implementation program of specified to achieve the HLW disposal. end state Because the rough description of LLWs from FR cycle are not different from those from LWR cycle, the existing or developing technologies for LLWs disposal are also applicable to the disposal of LLWs from FR cycle. Those technologies have reached almost “commercial level”.
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TABLE B.6. RESULTS OF EVALUATION OF RADIOACTIVE WASTE MANAGEMENT USING THE INPRO METHODOLOGY (CONTINUED)
(IN3.1.3) Resources (funding, Yes Availability of resources space, capacity, etc.) The scheme is established so that the available for achieving cost for the geological disposal will the end state compatible be funded by the electricity utilities. with the size and Regarding the disposal cost for growth rate of the LLWs, legislating of funds for some energy system part of wastes is in progress. The quantities of LLWs including the LLWs from FR cycle should be considered by comparing the results of the “Second Progress Report on Research and Development for TRU Waste Disposal in Japan” (IN3.1.4) Meet regulatory Yes Safety of the end state (long- standards of specific A safety case for HLW disposal was term expected dose to an Member State already developed in H12 based on a individual of the critical generic safety assessment. group) Safety cases for LLWs disposal were also developed in the “Second Progress Report on Research and Development for TRU Waste Disposal in Japan” and the Nuclear Safety Commission presented the safety evaluation method and target dose rate. Moreover, the disposal of some LLWs has already started. Therefore, it is possible to judge if the regulatory standards will be met. (IN3.1.5) As short as reasonably Yes Time to reach the end state practicable The staged implementation process is promulgated in "the Specified Radioactive Waste Final Disposal Act" for HLW disposal. The time necessary for each step has been estimated for HLWs. The disposal of short lived LLWs has already started and the measures for the long-lived LLWs (ex. TRU wastes) will be taken steadily according to the Japan’s National Nuclear Energy Plan by the ANRE (Agency for Natural Resources and Energy).
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TABLE B.6. RESULTS OF EVALUATION OF RADIOACTIVE WASTE MANAGEMENT USING THE INPRO METHODOLOGY (CONTINUED)
UR: User Requirement UR3.2: (Attribution of Waste Management Costs): The costs of managing all waste in the life cycle should be included in the estimated cost of energy from the INS, in such a way as to cover the accumulated liability at any stage of the life cycle. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations (IN3.2.1) Included Included Specific line item in the cost All the costs needed for waste estimate included management (treatment, storage, transport, and disposal) are included in the cost evaluation of the INS in Japan although it should be refined according to the progress of the design study. BP: Basic Principle BP4: (Waste optimization) Interactions and relationships among all waste generation and management steps shall be accounted for in the design of the INS, such that overall operational and long-term safety is optimized. UR: User Requirement UR4.1: (Waste Classification): The radioactive waste arising from the INS should be classified to facilitate waste management in all parts of the INS. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations (IN4.1.1) Scheme permits Yes Classification Scheme unambiguous, At present, the scheme for the waste practical segregation management of the INS in Japan is and measurement of considered the same as that of the waste arising existing LWRs and their fuel cycle in Japan. The LLWs from LWR cycle are segregated or the measurement technologies necessary for the segregation are under development based on the scheme including the regulatory standards. It is possible to segregate and measure the LLWs from FR cycle because the characteristics of the LLWs are not so different from the LLWs from LWR cycle.
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TABLE B.6. RESULTS OF EVALUATION OF RADIOACTIVE WASTE MANAGEMENT USING THE INPRO METHODOLOGY (CONTINUED)
UR: User Requirement UR4.2: (Pre-disposal Waste Management): Intermediate steps between generation of the waste and the end state should be taken as early as reasonably practicable. The design of the steps should ensure that all-important technical issues (e.g., heat removal, criticality control, confinement of radioactive material) are addressed. The processes should not inhibit or complicate the achievement of the end state. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations (IN4.2.1) As short as reasonably Yes Time to produce the waste practicable form specified for the end It is possible to follow the principle state of “As short as reasonably practicable”. Regarding LLWs, the detailed ways to process original LLWs into the waste forms specified for end states are under development. (IN4.2.2) Criteria as prescribed Yes Technical indicators such as by regulatory bodies of The Japan’s regulatory body will be criticality compliance, specific Member States expected to establish the criteria for emissions control, radiation the waste management. All the shielding, waste form criteria will be observed when the INSs are deployed. The regulatory standards are expected to be satisfied because the characteristics of the LLWs from FR cycle are not different from those from LWR cycle. (IN4.2.3) Complete chain of Yes Process descriptions that processes from It is possible to describe the entire encompass the entire waste generation to final end waste life cycle in general. life cycle state and sufficiently detailed to make evident the feasibility of all steps
B6. PROLIFERATION RESISTANCE
B6.1. Introduction Japan’s studies on the proliferation resistance area in the Joint Study were carried out based on the INPRO manual. Japan has started the FaCT project to research and develop, and demonstrate
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the commercialized FR (Fast Reactor) cycle systems for the future. Since the FR cycle which consists of the JSFR and its fuel cycle facilities was selected as the principal concept of the project, it was also selected as the reference system in this study. Most of the indicators were evaluated in consideration of the technical state of the art and the development goals towards future Safeguards/proliferation resistance/physical protection for advanced fuel cycles. B6.2. Evaluation results The proliferation resistance area of INPRO has one basic principle (BP) and five user requirements (UR ). The BP and URs are as follows: Basic principle BP: Proliferation resistance intrinsic features and extrinsic measures shall be implemented throughout the full life cycle for innovative nuclear energy systems to help ensure that INSs will continue to be an unattractive means to acquire fissile material for a nuclear weapons program. Both intrinsic features and extrinsic measures are essential, and neither shall be considered sufficient by itself. UR1 State commitments: States’ commitments, obligations and policies regarding non- proliferation and its implementation should be adequate to fulfill international standards in the non-proliferation regime. UR2 Attractiveness of NM and technology: The attractiveness of nuclear material (NM) and nuclear technology in an INS for a nuclear weapons program should be low. This includes the attractiveness of undeclared nuclear material that could credibly be produced or processed in the INS. UR3 Difficulty and detectability of diversion: The diversion of nuclear material (NM) should be reasonably difficult and detectable. Diversion includes the use of an INS facility for the production or processing of undeclared material. UR4 Multiple barriers: Innovative nuclear energy systems should incorporate multiple proliferation resistance features and measures UR5 Optimization of design: The combination of intrinsic features and extrinsic measures, compatible with other design considerations, should be optimized (in the design/engineering phase) to provide cost-efficient proliferation resistance. Table B.7 shows the detailed evaluation results of the proliferation resistance area in Japan’s studies. As a conclusion, international societies should reach an international standard (consensus) to attain an effective and acceptable level in proliferation resistance (PR) measures.
B6.3. Comments to the INPRO methodology
• Numerical evaluation scales illustrated in the PR manual should be changed to be more qualitative ones. • There is a difficulty to evaluate some indicators at the stage of project planning. • Evaluation scale “analysis is not yet done, but to be done” is strongly recommended at the concept design stage. • It is suggested that the acceptance limits should be changed to “sufficient according to international practice”.
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• It is suggested that the evaluation scales for intrinsic features should be changed, and not used as the acceptance limits.
TABLE B.7. RESULTS OF EVALUATION OF PROLIFERATION RESISTANCE
BP: Basic Principle Proliferation resistance intrinsic features and extrinsic measures shall be implemented throughout the full life cycle for innovative nuclear energy systems to ensure that INSs will continue to be an unattractive means to acquire fissile material for a nuclear weapons program. Both intrinsic features and extrinsic measures are essential, and neither shall be considered sufficient by itself. UR: User UR1: States’ commitments, obligations and policies regarding non- Requirement proliferation and its implementation should be adequate to fulfil international standards in the non-proliferation regime. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations IN1: States’ In accordance with [Acceptance] commitments, international standards Japan is party to NPT. In Japan, a CSA and obligations and policies Additional protocol concluded with IAEA established? and Export controls are in place. The legal framework of nuclear law including SSAC in Japan is adequately established in accordance with international standards. The national authority dealing with safeguards is established.
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TABLE B.7. RESULTS OF EVALUATION OF PROLIFERATION RESISTANCE (CONTINUED)
UR: User UR2: The attractiveness of nuclear material (NM) in an INS for a Requirement nuclear weapons program should be low. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations IN2.1: AL2.1: Attractiveness [comments] (Same as IN2.1 to IN2.4) NM quality based on NM On the basis of PR strategy* of FR Cycle characteristics considered in Japan, the fuel is unattractive for IN2.2: in design of INS and diversion and believed to be inherently PR NM quantity found acceptable low *No pure Pu in the cycle, but mixed based on expert judgment U/TRU fuel with low decontamination IN2.3: (containing FPs) NM form Numerical evaluation scales illustrated in the draft PR manual should be changed to IN2.4: Nuclear be more qualitative ones. technology
All EPs will be over Strong. (Design concept base)
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TABLE B.7. RESULTS OF EVALUATION OF PROLIFERATION RESISTANCE (CONTINUED)
UR: User UR3: The diversion of nuclear material should be reasonably difficult Requirement and detectable. (continued) CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations IN3.3: Detectability of AL3.3: Based on expert [Acceptance] NM judgment equal or better All EPs will be over Strong. (Design than existing facilities concept base)
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TABLE B.7. RESULTS OF EVALUATION OF PROLIFERATION RESISTANCE (CONTINUED)
UR: User UR4: Innovative nuclear energy systems should incorporate multiple Requirement proliferation resistance features and measures. CR: Criteria IN: Indicator AL: Acceptance Limit Evaluation results and Explanations IN4.1: The extent to AL4.1: All plausible [Acceptance] which the INS is acquisition paths are The intrinsic features will be included to covered by multiple (can be) covered by decrease the impact of safeguards intrinsic features and extrinsic measures on implementation and verification on the facility extrinsic measures the facility or State operations. level and by intrinsic features which are
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B7. INFRASTRUCTURE
B7.1. Introduction Japan’s infrastructure assessment was carried out based on the INPRO manual. The indicators of infrastructure were evaluated based on the progress of the FaCT project. The reference nuclear system of the FaCT project (JSFR) is a sodium-cooled fast reactor cycle system with a MOX fuel core. However, as most of the INPRO requirements of infrastructure are used to arrange the infrastructure for an INS, these requirements have more a function like a checklist. Japan has long experience with nuclear power generation, and many nuclear plants are operating now. Japan has also plans for further INS deployment. This experience of nuclear power deployment was reflected in the evaluation of the infrastructure. B7.2. Evaluation results INPRO infrastructure basic principle (BP1) states “Regional and international arrangements shall provide options that enable any country that so wishes to adopt INS for the supply of energy and related products without making an excessive investment in national infrastructure”. This basic principle has four user requirements: UR1.1 ‒ legal and institutional infrastructure, UR1.2 ‒ economic and industrial infrastructure, UR1.3 ‒ socio-political infrastructure, UR1.4 ‒ infrastructure of human resources. There are several indicators for each user requirement, and an indicator is evaluated by use of evaluation parameters and their acceptability. All indicators were checked with the evidence concerning acceptance. Table B.8 shows the detailed evaluation results of infrastructure in Japan’s studies. The completeness and adequacy of nuclear law is evaluated in UR1.1. Japan has experience with nuclear power generation. The legal framework of nuclear law in Japan is adequately established in accordance with international standards. Thus the evaluation result of UR1.1 in Japan’s studies was ‘acceptance’. Availability of credit lines, size of installation, support infrastructure, cost-benefit are evaluated in UR1.2. The results of a cost-benefit analysis in the FaCT project showed that the investment for R&D brings in good returns through the commercialized fast reactor cycle deployment. FaCT worked out a road map of the project of future JSFR, which was authorized by the government. FaCT specified installed capacities the nuclear power generation based on energy demand and supply prospect in the future and estimated the size of installations. In the design analysis of FaCT, technical feasibility and business applicability were confirmed by a joint study with domestic manufacturers. The domestic manufacturers have construction experience of many nuclear power plants. Thus the evaluation result of UR1.2 in Japan’s studies was ‘acceptance’. The national energy strategy, communications, and social acceptability are evaluated in UR1.3. Some related policies of Japan are established, which include the new national energy strategy, framework for nuclear energy policy, and an energy basic plan. A web page for support of education, which provides knowledge and information about the nuclear energy, is available as a communication tool. Public comments, public hearings, and the safety agreement, which are always linked to each other in the local prefecture, as well as risk communication, are published on the web to foster public acceptance. Thus, the evaluation result of UR1.3 was ‘acceptance’ based on such an action program of public acceptance. However, it should be noted that the effort to foster public acceptance and the level of public acceptance of a (new) nuclear installation are
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not always related. There are challenges to reduce public anxiety regarding radioactive substances and site selection; in order to accept a nuclear installation, realistically, though an effort for improvement of public acceptance is indispensable. Therefore, the effect of public acceptance should be confirmed in order to feedback continuously. Human resources and safety culture prevalence are evaluated in UR1.4. A learning opportunity about nuclear energy and training at public research organizations are provided in Japan. The human resource cultivation is advanced in cooperation with the local community and university. A qualification merit, education training and a technical expert training-up system are provided for the benefits to society. As part of these efforts from nuclear-related industries the nuclear safety network was established. The main activity is the dissemination of the safety culture of nuclear energy, a peer review among the members, support for communication, and education about nuclear safety. As part of the efforts of the government and atomic energy commission, a report of the ideal way of a nuclear safety culture is published, and a safety culture opinion exchange meeting is held. Internationally, Japan has joined WANO. Japan also participates in the activity of the forum for nuclear cooperation in Asia (FNCA) and is exchanging views and communications about nuclear safety culture and human resources development. Thus, the evaluation result of UR1.4 was ‘acceptance’ based on such an action of public acceptance. As Japan has long experience with nuclear power generation, the evaluation results of infrastructure in Japan become “all acceptances”. Difficulties are anticipated to evaluate some indicators for a country which plans to construct an infrastructure from now, because the appraisable indicators both at the planning stage and the construction phase are difficult to obtain. However, for a country which plans to construct an infrastructure from now, these indicators are useful and valid as a check list. B7.3. Comments to the INPRO methodology As a result of the evaluation of infrastructure, there are comments to the INPRO methodology as follows. • For the country which plans to construct an infrastructure from now, these indicators are valid as a check list. • Most of indicators evaluate the nuclear development conditions in a country rather than the difference of INSs. • There is a difficulty to evaluate some indicators at the stage of project planning. (ex. IN1.1.1, IN1.1.2, IN1.2.1, IN1.3.1, IN1.4.2, IN1.4.3) • IN 1.3.1, IN 1.3.2: It is suggested that the acceptance limits for these indicators are “sufficient according to international practice” not “sufficient according to the best international practice”. • IN 1.4.3: This acceptance limits require cost estimation. However it is difficult to evaluate “Benefits to society” quantitatively. Only qualitative judgment is possible.
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY
BP: BP1: Regional and international arrangements shall provide options that enable any Basic country that so wishes to adopt INS for the supply of energy and related products Principle without making an excessive investment in national infrastructure. UR: User Requirement UR1.1: Legal and institutional infrastructure CR: Criteria IN: Indicator EP: Evaluation AL: Evaluation results Parameters Acceptance Limit IN1.1.1 EP1: In accordance [Acceptance] Legal framework completeness of with Japan has experience with the established nuclear law international nuclear power generation. The EP2: adequacy of standards legal framework of nuclear law in nuclear law Japan is adequately established in EP3: scope of accordance with international international standards. legislation in the
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY (CONTINUED)
UR: User Requirement UR1.2: Economic and industrial infrastructure CR: Criteria IN: Indicator EP: Evaluation AL: Evaluation results Parameters Acceptance Limit IN1.2.1 EP1: credit lines Sufficient to [Acceptance] Availability of credit for infrastructure cover the The results of the cost-benefit lines provided by project analysis showed that the industry and investment for the R&D brings in operator good returns through the EP2: credit lines commercialized FR cycle for infrastructure deployment. The result of provided by investment risk which was government evaluated for the investment amount and the payout time, was manageable.
Financing facility like government subvention or federal aid for nuclear power.
IN1.2.2 (Not EPs Adequate to [Acceptance] Demand for and price identified) enable ROI Electricity generation cost (UR of energy products 1.1) for SFR in economics evaluation is adequate to enable ROI.
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY (CONTINUED)
IN1.2.3 EP1: energy Matches local [Acceptance] Size of installation system expansion needs The FaCT worked out the road plan map of the project of future EP2: size of JSFR, which was authorized by nuclear fuel cycle the government. The FaCT facilities (other specifies that a nuclear power than NPP) generation installed capacities from the energy demand and supply prospect for the future and assumes the size of installation.
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY (CONTINUED)
UR: User Requirement UR 1.3: Socio-political infrastructure CR： Criteria IN: Indicator EP: Evaluation AL: Evaluation results Parameters Acceptance Limit IN1.3.1 EP1: existence of Sufficient [Acceptance] Information to public a national energy according to Some related policies of Japan (e.g., appropriate national energy strategies) policy the best are established. EP2: international
The operation information in the nuclear facilities can be obtained from the web page of each electric power company.
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY (CONTINUED)
EP4: addressing [Acceptance] of concerns raised Common understanding of the by the public nuclear among the public regarding nuclear administration, the nuclear installations industries and the local residents EP5: use of is developed by the risk communication communication approach. experts to match information to To improve the skill of the audience communication expert, there are education and training as well as a qualification system. (see IN1.4.1, IN1.4.3) IN1.3.2 EP1: appropriate Sufficient [Acceptance] Participation of the decision making according to In the policy formation process, Atomic Energy Commission invites the public public in the decision process the best to offer their opinions (the public making process (to i) Accessibility to international comment), and incorporates a wide foster public resources practice range of views from the general public. acceptance) ii) Task definition As the mechanism of the national iii) Structured consensus building up the nuclear politics, the citizen participation round- decision-making table conference to deepen mutual process understanding is held in various places. iv) Cost As for the installation of the nuclear effectiveness facilities, a public hearing is held to EP2: obtain the opinion of the residents and make allowance for it. Acceptability of Safety Agreement is always linked the results among the local prefecture, the towns i) Sample and villages, the neighbor towns and representative villages and the nuclear industries, and ii) Independence the public has the right of various demands to the nuclear industries. of the Safety Agreement is always linked participation among the local prefecture, the towns process and villages, the neighbor towns and iii) Early villages and the nuclear industries, and involvement the public has the right of various iv) Influence of demands to the nuclear industries. results on policy
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY (CONTINUED)
IN1.3.3 EP1: Sufficient to [Acceptance] Long term confirmation of enable ROI The policies such as “New commitment long-term National Energy Strategy " and commitment by "Framework for Nuclear Energy government Policy " in the government are prepared and aimed at mid and long term goals. IN1.3.4 EP1: surveys Sufficient to [Acceptance] Public acceptance of about public ensure there is Atomic Energy Commission nuclear power acceptance negligible invites the public comment, and performed on a political or incorporates a wide range of regular basis policy risk to views from the general public. EP2: adequacy of the investment As for the installation of the survey nuclear facilities, a public EP3: acceptable hearing is held to obtain the result of survey opinion of the residents and make allowance for it. The public administration, the nuclear industries and the local resident are advancing mutual understanding by the risk communication. The nuclear industry carries out the public hearing and public relations using the dialogue, the experience and the media to promote mutual understanding with the people.
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY (CONTINUED)
UR: User Requirement UR1.4: Infrastructure of human resources CR: Criteria IN: Indicator EP: Evaluation AL: Evaluation results Parameters Acceptance Limit IN1.4.1 EP1: educational Sufficient [Acceptance] Availability of human and training according to The learning opportunity about the nuclear energy and the training at the resources system for international public research organization are manpower needed experience provided. in NP projects The human resource cultivation is EP2: professional advanced in cooperation with the local perspectives in community and university.
the nuclear power
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TABLE B.8. RESULTS OF THE INFRASTRUCTURE CRITERIA EVALUATION USING THE INPRO METHODOLOGY (CONTINUED)
IN1.4.3 EP1: skills BTS > cost [Acceptance] Benefits to society of improvement of necessary to The qualification merit, the the INS (human human resources establish and education training, and the resources) provided by maintain the technical expert training-up industry required system are provided for the EP2: study on expertise benefits to society. (see. IN1.4.1) governmental investment in
B8. SCENARIO ANALYSIS
B8.1. Introduction Based on the boundary conditions for INS evaluation in the Joint Study, we aim at a maximum scale of fast reactor (FR) deployment, natural uranium saving benefits and so on, when the JSFR (main concept selected in the FaCT project; sodium-cooled FR with MOX fuel) is deployed in the world. Timing of FR deployment in the key countries is expected to be by 2020−2035 (India: 2020, Russian Federation: 2020−2025, France: around 2030, China: 2030−2035, Japan: around 2050). B8.2. Assumptions The nuclear power generation capacity adopted in this scenario analysis is shown in Fig. B.5. Nuclear power generation capacity was calculated based on the nuclear power electric capacity in the IPCC SRES B2 scenario and an assumed load factor (90 %). Main assumptions for characteristic data of reactor systems and fuel cycle systems are shown in Table B.8 and B.9. The fuel burn-ups of LWR and FR are assumed to be 45−60 GWd/t and 55−115 GWd/t (average burn-ups), respectively. The breeding ratios of FR are about 1.03 and about 1.20, and there will be a switchover to a low breeding type core according to the Pu balance. Lifetime for every type of reactor is assumed to be 60 years. The ex-core time periods have been assumed to be 4 years for LWR cycle and 5 years for FR cycle. The tails assay in enrichment plant is assumed to be 0.25 %.
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The list of analysis cases is shown in Table B.11. B8.3. Evaluation results The conclusion from the analysis of a global scenario is as follows. (1) In IPCC SRES-B2 scenario, cumulative natural uranium demands will exceed the total conventional uranium resources of 14.8 million ton-U (OECD/NEA-IAEA, Uranium 2005), even if FRs are deployed in 2020 (Fig. B.9−B.12); (2) Under constrained conditions of total conventional uranium resources, it is difficult to supply electricity assuming the IPCC SRES-B2 scenario. (Fig. B.14−B.18); (3) Shortening the duration of the ex-core time period and decreasing tails assay concentration will lead to a decrease of cumulative natural uranium demands.(Table B.12) (4) However, the shortage of nuclear power generation capacity will be suppressed the most when ex-core time period is 3 years (Case-B3).
120,000 IPCC-A1C IPCC-A2
IPCC-B1 ※1 100,000 IPCC-B2
IPCC SRES-B2 IIASA-A1 80,000 IIASA-A2 IIASA-A3 60,000 IIASA-B IIASA-C1 40,000 IIASA-C2 IEA-Alternative Scenario 20,000 IEA-Reference Scenario generation power Nuclear (TWh) IAEA-Low 0 IAEA-High 1990 2000 2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Year
FIG. B.5. Typical perspectives for nuclear power generation. Note: IPCC= Intergovernmental Panel on Climate Change; IIASA=International Institute for Applied System Analysis; WEC=World Energy Council; Results produced by the MESSAGE computer code.
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Assumption of Load Factor 90% 6,000 Nuclear capacity after 2100 is
5,000 constant HWR-U(10%)
4,000
3,000 APWR(70%)
2,000 PWR Nuclear Capacity (GWe) 1,000 BWR ABWR(20%) 0 1950 2000 2050 2100 2150 2200 Year
FIG. B.6. World nuclear power capacity (No FR deployment).
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TABLE B.9. ASSUMPTIONS OF NUCLEAR POWER REACTOR SYSTEM CHARACTERISTIC DATA
Life time 60 years for all type of reactors Reactor type BWR Burn-up 45 GWd/t for reactors that will be deployed by 2019 ABWR Burn-up 60 GWd/t for reactors that will be deployed by 2020 PWR Burn-up 49 GWd/t for reactors that will be deployed by 2019 APWR Burn-up 60 GWd/t for reactors that will be deployed by 2020 HWR-U Burn-up 8.3 GWd/t natural uranium fuel for CANDU FR High breeding ratio 1.20, average burn-up 55 GWd/t. sodium cooled , MOX fuel Low breeding ratio 1.03, average burn-up 115 GWd/t. sodium cooled , MOX fuel Deployment BWR&ABWR 20 %of all capacity except FR ratio PWR&APWR 70 %of all capacity except FR HWR-U 10 %of all capacity except FR
TABLE B.10. ASSUMPTIONS OF NUCLEAR FUEL CYCLE SYSTEM CHARACTERISTIC DATA
Ex-core time LWR 4 years period ( Cooling time 3years, Reprocessing 0.5 year, Fabrication & transportation 0.5 year) FR 5 years (Cooling time 4years, Reprocessing 0.5 year, Fabrication 0.5 year) (Cooling time 4years, Reprocessing 0.5 year, Fabrication 0.5 year) Enrichment plant Capacity is not limited. Fuel fabrication plant Capacity is not limited. Reprocessing LWR-UO2 ～2009：4,100 ｔon-HM/year Plant 2010～justTable before 8.3 FR development List of analysis： 4,900 caseton HM/year After FR deployment： The reprocessing plant capacity is gradually increased in accordance with amount of spent fuels discharged from FR.
LWR-MOX They will be processed in the FR reprocessing plant in 20-40 years. HWR Long-term storage of spent fuels FR Reprocessing of all spent fuels Loss factor LWR, Enrichment 0%、Conversion0.5%、 Fabrication 0.1%、 Reprocessing 0.5% HWR FR Fabrication 0.1%、 Reprocessing 0.1%
U-235 enrichment 0.25%
Other The uranium recovered from spent fuels is re-enriched
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TABLE B.11. LIST OF ANALYSIS CASES
FR deployment Tails assay Ex-core time period Case In 2020 In 2030 0.25% 0.20% 5 years*1 3 years*2
LWR once-through ― （A1） ― ○ Deployment in 2030 Resources （A2） ○ ○ ○ free Deployment in 2020 ● ○ ○ （A3） Uranium saving（A4） ○ ● ○ LWR once-through ○ ○ （B1） Base case （B2） Resources ○ ○ ○ 3 Ex-core Time 3 years restriction* ○ ○ ● （B3） Uranium saving（B4） ○ ● ○ *1： 5 years (Cooling time 4years, Reprocessing 0.5 year, Fabrication 0.5 year) Note *2： 3 years (Cooling time 2years, Reprocessing 0.5 year, Fabrication 0.5 year) *3： Limitation of natural uranium 14.8 million ton U after 2005
Case-A1 7,000 Total 6,000
5,000 HWR 4,000 LWR 3,000 Pu recycling 2,000 in LWR ALWR Nuclear Capacity (GWe) 1,000
0 2000 2020 2040 2060 2080 2100 Year
FIG. B.7.World nuclear capacity (Case-A1： LWR once-through).
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*1 OECD/NEA-IAEA. Uranium 2005:Resources, Production and Demand(2006) 50
40 39
30
20 Total Conventional Uranium （ ）*1 (million tonU) (million Resources 14.8 million tonU
10 8 Accumulative uranium demands 0 2000 2020 2040 2060 2080 2100 Year
FIG. B.8. World cumulative natural uranium demands after 2005(Case-A1： LWR once-through).
Case-A2&A4 7,000 Total 6,000
5,000 HWR 4,000 FR (B.R.1.2) 3,000 Pu recycling in LWR FR 2,000 (B.R.1.03)
Nuclear Capacity (GWe) 1,000 Adv. LWR LWR 0 2000 2050 2100 2150 2200 Year
FIG. B.9. World nuclear capacity (Case-A2 and A4: FR deployment in 2030).
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Case-A3 7,000 Total 6,000
5,000 HWR 4,000 FR (B.R.1.2) 3,000 Pu recycling in LWR FR 2,000 (B.R.1.03)
Nuclear Capacity (GWe) 1,000 Adv. LWR LWR 0 2000 2050 2100 2150 2200 Year
FIG. B.10. World nuclear capacity (Case-A3： FR deployment in 2020).
Case-A2: FR introduction in 2030 30 (Tails assay 0.25%) 26.6 25 25.4 Case-A3: FR introduction in 2020 24.4 (Tails assay 0.25%) 20 Case-A4: FR deployment in 2030 (Tails assay 0.20%) 15
10 (million tonU) (million Total Conventional Uranium Resources after 2005 5 （14.8 million tonU） Accumulative uranium demands demands uranium Accumulative 0 2000 2050 2100 2150 2200 Year
FIG. B.11. World cumulative natural uranium demands after 2005.
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Case-A2 50,000 LWR reprocessing plants capacity
40,000
30,000
20,000
Amount of processed 10,000 LWR(UO2) SF Reprocessing Capacity (ton/year) Capacity Reprocessing 2
0 2000 2050 2100 2150 2200 Year
Fig. B.12. LWR reprocessing plants capacity in the world (Case-A2: FR deployment in 2030).
Case-A2 50,000 FR reprocessing plants capacity
40,000
30,000 LWR（MOX) SF 20,000 Amount of processed FR SF 10,000
Reprocessing Capacity (ton/year) 0 2000 2050 2100 2150 2200 Year
FIG. B.13. FR reprocessing plants capacity in the world (Case-A2: FR deployment in 2030).
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Case-B1 7,000 IPCC SRES-B2 6,000
5,000
4,000 HWR 3,000 Pu recycling Adv. LWR in LWR 2,000
Nuclear CapacityNuclear (GWe) 1,000 LWR 0 2000 2050 2100 2150 2200 Year
FIG. B.14. World nuclear capacity (Case-B1: LWR once-through).
Case-B2 7,000 Shortage of 1,630GWe IPCC SRES-B2 6,000
5,000
4,000 city (GWe) FR 3,000 (B.R.1.2) Pu recycling in LWR 2,000 HWR Adv. LWR Nuclear Capa 1,000 FR LWR (B.R.1.03) 0 2000 2050 2100 2150 2200 Year
FIG. B.15. World nuclear capacity (Case-B2： Base case).
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Case-B3 7,000 Maximum shortage of IPCC SRES-B2 360GWe 6,000
5,000
4,000 FR 3,000 (B.R.1.2) Pu recycling in LWR FR 2,000 in LWR LWR (B.R.1.03) HWR Nuclear Capacity (GWe) 1,000
ALWR 0 2000 2050 2100 2150 2200 Year
FIG. B.16. World nuclear capacity (Case-B3： Ex-core time period 3years).
Case-B4 7,000 Maximum shortage of IPCC SRES-B2 1,360GWe 6,000
5,000
4,000
3,000 Pu recycling FR in LWR (B.R.1.2) 2,000 LWR HWR Nuclear Capacity (GWe) 1,000 FR ALWR (B.R.1.03) 0 2000 2050 2100 2150 2200 Year
FIG. B.17. World nuclear capacity (Case-B4： Tails assay 0.2 %).
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Limit of natural uranium 14.8 million ton-U after 2005 30
25
20 Total Conventional Uranium Resources （14.8 million tonU） 15 14.7 14.3 Once- 10 through(B1) (million tonU) FR deployment(B2-B4)
5
Accumulative uranium demands 0 2000 2050 2100 2150 2200
Year
FIG. B.18. World cumulative natural uranium demand after 2005 (Case B1-B4).
TABLE B.12. SUMMARY OF GLOBAL SCENARIO ANALYSIS
Cumulative natural uranium The year when The year when demand (million ton U) Nuclear shift to FR will nuclear Case capacity in 2100 finish generation will at 2100 maximum (GWe) reach SRES-B2 (year) case LWR once-through （A1） 39.1 － 5,000 － － Introduction in 2030 19.1 26.6 5,000 2160 － Resource （A2） free Introduction in 2020 （A3） 18.2 25.4 5,000 2160 － Uranium saving （A4） 17.4 24.4 5,000 2160 － LWR once-through （B1） 14.3 － 290 － － Resource Base case （B2） 13.7 14.4 3,371 2121 2132 restriction ※ Ex-core Time 3years（B3） 13.5 14.7 4,641 2128 2106 Uranium saving （B4） 13.6 14.7 3,642 2125 2129
*) In the year when all LWRs will be replaced by FR
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B9. CONCLUSIONS
As mentioned above, the evaluation result of the JSFR and its fuel cycle facilities almost achieved the acceptance level of the INPRO methodology. In addition, the establishment of innovative technologies allows us to improve the requirement and criteria of the INPRO. In the process of evaluation by using the INPRO methodology, Japan identified some findings and comments on the INPRO methodology for each area (i.e., Economics, Environment, Safety, Waste Management, Proliferation Resistance, Infrastructure, and Scenario Analysis). In general, the INPRO methodology is useful for an INS to develop and assess the conceptual design in an early stage. However, we found it difficult to use some INPRO indicators and evaluate the target by these indicators because of each nation’s status, wide-ranging scope, unclear criteria, difficulty of quantitative analysis, and so on. These outstanding issues should be discussed further among the interested countries in order to build a consensus. And then, it could be assumed that improvements will be reflected into the existing INPRO methodology in the future.
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ABBREVIATIONS
AECL Atomic Energy of Canada Limited ADS accelerator driven system AGR advanced gas reactor AHWR advanced heavy water reactor AL acceptance limit (INPRO) ALARP as low as reasonably practical, social and economic factors taken into account BP basic principle (INPRO) BR breeding ratio CDA core disruptive accident CEA Atomic Energy Commission (France) CEFR China Experimental Fast Reactor CIAE China Institute of Atomic Energy CNFC-FR closed nuclear fuel cycle with fast reactors CR criterion (INPRO) CP collaborative project (INPRO) DBE design basis earthquake DHRS decay heat removal system DTV desired target value (INPRO) DU depleted uranium EFR European fast reactor EPR European pressurized reactor FC fuel cycle FCF fuel cycle facility FOAK first-of-a-kind FP fission products FBR fast breeder reactor FR fast reactor FSA fuel subassembly GCC gas combined cycle GFR gas cooled fast reactor
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GHG greenhouse gas GIF Generation-IV International Forum GW(e) Gigawatt electrical HEU highly enriched uranium HLMC heavy liquid metal coolants HLW high level radioactive waste HM heavy metal HTGR high temperature gas reactor HTR high temperature reactor HWR heavy water reactor I&C instrumentation and control ICG international coordinating group in INPRO IDC interest during construction IGCAR Indira Gandhi Centre for Atomic Research IIASA International Institute for Applied System Analysis IN indicator (INPRO) INPRO International Project on Innovative Nuclear Reactors and Fuel Cycles INS innovative nuclear energy system (INPRO) IPCC Intergovernmental Panel on Climate Change IPPE Institute for Physics and Power Engineering (Russian Federation) IRR internal rate of return ISI in-service inspection JAEA Japan Atomic Energy Agency KAERI Korean Atomic Energy Research Institute KI key indicator (INPRO) LCA life cycle assessment LCI life cycle inventory LDC levelized discounted cost LEU low enriched uranium LHR linear heat generation rate LMR liquid metal cooled reactor LWR light water cooled reactor MA minor actinides
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MCA&P material control, accounting and protection MFA material flow assessment MNFC multilateral nuclear fuel cycle MOX mixed oxide uranium-plutonium fuel MW(th) megawatt thermal MS Member State (IAEA) NNEGC National Nuclear Energy Generating Company (Ukraine) NPP nuclear power plant NPV net present value NPT Non-Proliferation Treaty NOAK Nth of a kind NS nuclear system ODS oxide dispersion strengthened steel O&M operation and maintenance OTFC one-through (open) fuel cycle P&T partitioning and transmutation PHWR pressurized heavy water reactor PIE post irradiation examination PR proliferation resistance (INPRO) PSA probabilistic safety analysis (assessment) PUREX plutonium and uranium recovery by extraction PWR pressurized water reactor RBMK graphite moderated fuel channel reactor R&D (RD&D) research and development (research, development and demonstration) ROI return on investment RRI relative risk index (INPRO) SF (SNF) spent fuel (spent nuclear fuel) SFR sodium cooled fast reactor SRES special report on emission scenarios STC scientific and technical committee of the Joint Study SWU separation work units TBP tri-butyl phosphate TOR terms of reference (IAEA)
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TR thermal reactor TRU transuranic elements ULOF unprotected loss of flow accident ULOHS unprotected loss of heat sink accident UR user requirement (INPRO) UTOP unprotected transient overpower accident VVER (WWER) water cooled, water moderated power reactor
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CONTRIBUTORS TO THE JOINT STUDY
The Joint Study report was developed by the members of Scientific and Technical Committee of the Joint Study nominated to the Committee by the official bodies of the participating countries: G. Dyck (AECL, Canada), Y.I. Kim (KAERI, Republic of Korea), N. Kukharchuk (UEA, Ukraine), X. Mi (CIAE, China), R. Nakai (JAEA, Japan), B. Raj (IGCAR, India), A. Vasile (CEA, France), S. Yugay, and V. Kagramanyan (IPPE, Russian Federation). In the following an alphabetic list of all experts contributing to the Joint Study is provided. Afanasyev, A. National Nuclear Energy Generating Company (UEA), Ukraine Aso, R. Japan Atomic Energy Agency (JAEA), Japan Busurin, Y. International Atomic Energy Agency Bychkov, A. Research Institute of Atomic Reactors (RIAR), Russian Federation Challapandi, P. Indira Gandhi Centre for Atomic Research, (IGCAR), India Dyck, G. Atomic Energy Canada Limited (AECL), Canada Fesenko, G. Obninsk State Technical University for Nuclear Enegineering (IATE), Russian Federation Grenèche, D. AREVA, France Hongyi, Y. China Institute of Atomic Energy (CIAE), China Indira, R. Indira Gandhi Centre for Atomic Research, (IGCAR), India Kagramanyan, V. Institute for physics and power engineering (IPPE), Russian Federation Kawasaki, H. Japan Atomic Energy Agency (JAEA), Japan Kim, Y. Korean Atomic Energy Research Institute (KAERI), Republic of Korea Kormilitsyn, M. Research Institute of Atomic Reactors (RIAR), Russian Federation Kukharchuk, N. National Nuclear Energy Generating Company (UEA), Ukraine Kurisaka, K. Japan Atomic Energy Agency (JAEA), Japan Leclere, J. Commissariat a l’Energie Atomic (CEA), France Loaec, C. Commissariat a l’Energie Atomic (CEA), France Mi, X. China Institute of Atomic Energy (CIAE), China) Mohanakrishnan, P. Indira Gandhi Centre for Atomic Research, (IGCAR), India Moulin, V. Commissariat a l’Energie Atomic (CEA), France Nagarajan, K. Indira Gandhi Centre for Atomic Research, (IGCAR), India Nakai, R. Japan Atomic Energy Agency (JAEA), Japan Ohtaki , A. Japan Atomic Energy Agency (JAEA), Japan Ono, K. Japan Atomic Energy Agency (JAEA), Japan
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Peide, Z. China Institute of Atomic Energy (CIAE), China Raj, B. Indira Gandhi Centre for Atomic Research, (IGCAR), India Sato, K. Japan Atomic Energy Agency (JAEA), Japan Shiotani, H. Japan Atomic Energy Agency (JAEA), Japan Subbotin, S. Kurchatov Institute (KI), Russian Federation Tsibulskiy, V. Kurchatov Institute (KI), Russian Federation Usanov, V. International Atomic Energy Agency Vasile, A. Commissariat a l’Energie Atomic (CEA), France Vasudeva Rao, P.R. Indira Gandhi Centre for Atomic Research, (IGCAR), India Yugay, S. Institute for Physics and Power Engineering (IPPE), Russian Federation Zongmao, G. China Institute of Atomic Energy (CIAE), China
Members of the Joint Study Scientific and Technical Committee of the Joint Study express their gratitude for valuable contributions to the study and discussions during the consultancy meetings to their colleagues from the Joint Study National Working Groups: Z. Peide, G. Zhongmao, Z. Donghui, Z. Weifang, Y. Guoan, Y. Hong, Y. Hongyi (CIAE, China); F.L. Linet, V. Moulin, G.L. Fiorini, J.Cazalet (CEA, France), D.Grenèche (AREVA, France); P.R. Vasudeva Rao, P. Chellapandi, R. Indira, R. Natarajan, Prabhat Kumar, V. Rajagopal, R. Kesavamurthy, P.K. Wattal, P. Puthiavinayagam (IGCAR, India); K. Ono, K. Sato, K. Kurisaka, H. Shiotani, R. Aso, H. Kawasaki, A. Ohtaki (JAEA, Japan); C.Y. Lim, Y.B. Lee, E.H. Kim, J.W. Choi, W.S. Chung, M.S. Young (KAERI, Republic of Korea); A. Chebeskov, I. Kuznetsov (IPPE, Russian Federation), P. Alekseev (KI, Russian Federation) A. Bychkov (RIAR, Russian Federation), P. Poluektov (VNIINM, Russian Federation).
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CONTRIBUTORS TO THE JOINT STUDY
The Joint Study report was developed by the members of Scientific and Technical Committee of the Joint Study nominated to the Committee by the official bodies of the participating countries: G. Dyck (AECL, Canada), Y.I. Kim (KAERI, Republic of Korea), N. Kukharchuk (UEA, Ukraine), X. Mi (CIAE, China), R. Nakai (JAEA, Japan), B. Raj (IGCAR, India), A. Vasile (CEA, France), S. Yugay, and V. Kagramanyan (IPPE, Russian Federation). In the following an alphabetic list of all experts contributing to the Joint Study is provided. Afanasyev, A. National Nuclear Energy Generating Company (UEA), Ukraine Aso, R. Japan Atomic Energy Agency (JAEA), Japan Busurin, Y. International Atomic Energy Agency Bychkov, A. Research Institute of Atomic Reactors (RIAR), Russian Federation Challapandi, P. Indira Gandhi Centre for Atomic Research, (IGCAR), India Dyck, G. Atomic Energy Canada Limited (AECL), Canada Fesenko, G. Obninsk State Technical University for Nuclear Enegineering (IATE), Russian Federation Grenèche, D. AREVA, France Hongyi, Y. China Institute of Atomic Energy (CIAE), China Indira, R. Indira Gandhi Centre for Atomic Research, (IGCAR), India Kagramanyan, V. Institute for physics and power engineering (IPPE), Russian Federation Kawasaki, H. Japan Atomic Energy Agency (JAEA), Japan Kim, Y. Korean Atomic Energy Research Institute (KAERI), Republic of Korea Kormilitsyn, M. Research Institute of Atomic Reactors (RIAR), Russian Federation Kukharchuk, N. National Nuclear Energy Generating Company (UEA), Ukraine Kurisaka, K. Japan Atomic Energy Agency (JAEA), Japan Leclere, J. Commissariat a l’Energie Atomic (CEA), France Loaec, C. Commissariat a l’Energie Atomic (CEA), France Mi, X. China Institute of Atomic Energy (CIAE), China) Mohanakrishnan, P. Indira Gandhi Centre for Atomic Research, (IGCAR), India Moulin, V. Commissariat a l’Energie Atomic (CEA), France Nagarajan, K. Indira Gandhi Centre for Atomic Research, (IGCAR), India Nakai, R. Japan Atomic Energy Agency (JAEA), Japan Ohtaki , A. Japan Atomic Energy Agency (JAEA), Japan Ono, K. Japan Atomic Energy Agency (JAEA), Japan
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