SODIUM-COOLED FAST REACTORS

Transcription

1 SODIUM-COOLED FAST REACTORS Lehrstuhl für Nukleartechnik - Technische Universität München Boltzmannstr Garching

2 1 SODIUM-COOLED FAST REACTORS 1.1 CONCEPT DEVELOPMENT The Generation IV Roadmap selected the sodium fast reactor (SFR) concept as one of the six technologies for further development under Generation IV. GIF has established a Steering Committee for the development of SFR, with participation of Japan, Republic or Korea, France, United Kingdom, United States, and Euratom. There are active and large development programs in Japan and Korea to develop an advanced SFR. Today construction of sodium-cooled fast reactors is proceeding in India, China, and Russia. Japan is working to bring on line again their prototype plant MONJU. Current development of advanced sodium-cooled reactors is focused on resolving remaining technology challenges for meeting the goals of the Generation IV initiative. The focus is on the development of advanced fuels with minor actinides, demonstration of advanced fuel recycle technologies, development of advanced inservice inspection technologies and improvement in the economics of the system. Satisfactory resolution of these challenges will demonstrate the ability of sodiumcooled reactors for meeting future energy demands and satisfying the Generation IV goals of sustainability, security, safety, and economics. For the use in fast reactors, Sodium offers a number of advantages. It has superior heat transport characteristics, excellent neutronics characteristics, is compatible with steel structural materials, and is relatively cheap. Sodium is a low density, high heat capacity material, which also results in low pumping power requirements compared to heavy liquid metals. Sodium coolant allows for a compact system design, even when allowing for passive core cooling. Moreover, the temperature range of sodium-cooled reactors allows operation at atmospheric pressure, minimizing the concerns of loss of coolant events that typically drive the safety design of water-cooled reactors. On the other hand, sodium is highly reactive with oxidants, including water and air, it has a relatively low boiling point, and it is activated by neutron capture (the activation product 24 Na is short-lived with 15 h half-life and decays in the stable 24 Mg that is 1

3 stable). The chemical activity of sodium requires that it is to be maintained under inert atmosphere. This resulted in the complication of the system design to reach the safety goals. It is possible to consult a complete database of information concerning Liquid Metal Cooled Fast Reactors at the following internet address: 2

4 1.2 TECHNICAL ASPECTS Different concepts for the design of a sodium cooled fast reactor are under study and they are divided essentially in two different groups: large monolithic plants medium sized plants For the first type reactors the configuration of the primary is loop type. The most significant project regarding this design type is a Japanese project: the JSFR, with an electric output of 1500 MW using mixed oxide fuel. In the second type reactors the configuration of the primary circuit is pool type. The most significant project regarding this design type is a Korean project: the Kalimer, with an electric output of 600 MW using metallic fuel. To be noticed is the recent idea that is gaining interest of small modular reactors (10 to 50 MWth) to be installed in remote locations with a small power need Loop type design: the JSFR The thermal power output of this commercial size reactor will be approximately 3750 MW with an efficiency of approximated 42%. The fuel used for the power generation is Pu-U mixed oxide. The reactor (Figure 1.1) is cooled by the sodium flowing in two closed loops. Figure 1.1: Scheme of the JSFR 3

5 The primary sodium loop operates near the atmospheric pressure and the heat generated in the nuclear fuel is transferred to the liquid metal that enters the core with a temperature of 668 K and is leaving it reaching a temperature of 823 K. In the JSFR, a fully natural circulation decay heat removal system will adopted for the purpose of improving the system reliability by not using active components, such as pumps and blowers. To be noticed is the reduction of the bundle pressure drop to 0.2 MPa so that the capability of removing the heat produced by the core in case of accident through natural circulation is enhanced. Sodium is very reactive and oxidizes quickly exothermal in water and air. For safety reasons a boundary between water and the sodium flowing in the primary loop is present and it is called the Intermediate Heat Exchanger (IHX). Then, trough the IHX, the heat is transferred to the second sodium loop called the Intermediate Heat Transport System (IHTS) in which sodium, moved by pumps, transports the heat to the steam generators and enters at a temperature of 608 K and reaches a temperature of 793 K. In the secondary loop the water is evaporated in the steam generators to a temperature of 768 K and a pressure of approximated 16.7 MPa departing from a temperature of 513 K. The mass flow rate of the feed water is approximated kg/s. The plant operating cycle is about 18 months, the burn-up is 150 GWd/t and the breeding ratio is approximated 1.2. The plant specification and the overall core design parameters of the JSFR are visible in Table 1.1 and Table

6 Table 1.1: Plant specification for the JSFR design Table 1.2: Core design parameters for the JSFR In Table 1.3 are reported general information on Administrative and responsible authorities of the JSFR. Table 1.3: JSFR general information (Reference - Fast Reactor Database) Location, postal address of station Owner Operator Designer Manufacturer or chief contractor Licensing authority Dates of major events - not determined not determined not determined not determined JNC/JAPC/ Mitsubishi (Hitachi/Toshiba/Fuji/Kawasaki) not determined 5

7 1.2.2 Pool type design: KALIMER The thermal power output of this pool type reactor will be approximately 1520 MW with an efficiency of approximated 39.4% much higher than that of the other nuclear power plants currently under operation in Korea. This high efficiency would enable to utilize the nuclear fuel more effectively and mitigate the thermal load on environment. The fuel used for the power generation is U-TRU- 10 Zr, a metallic fuel. The Nuclear Steam Supply System of the KALIMER-600 (Figure 1.2) consists of different subsystems configured in series. In the pool-type Primary Heat Transport System (PHTS) situated in a large sodium pool, the coolant enters at a temperature of 663 K and exits at a temperature of 818 K. The Intermediate Heat Transport System (IHTS) consists of two identical loops with the effect of contributing to the flexibility of plant operation and increasing the reliability of the decay heat removal by the normal procedures. The coolant enters here at a temperature of 594 K and reaches a temperature of 799 K. In the Steam Generation System (SGS) the steam is produced at a temperature of 776 K and a pressure of 16.5 MPa with a flow rate of 663 kg/s. The steam generator is a helical coil, vertically-oriented shell-and-tube type heat exchanger. A countercurrent flow is formed by the sodium flow through the shell side and water/steam through the tube side. The space inside the reactor vessel is thermally divided into two regions, hot region and cold region. The reactor vessel for KALIMER-600 contacts with the cold sodium at 663 K. Therefore, the reactor vessel can maintain its structural integrity beneficially for a 60 years design life time. The plant operating cycle is about 18 months, the burn-up is 66 GWd/t and the breeding ratio is approximately 1.0. Furthermore, the KALIMER is not loaded with any blanket assembly and an excessive plutonium production is minimized by keeping the core conversion ratio approximately equal to one. 6

8 Figure 1.2: Scheme of the primary of the KALIMER-600 In Figure 1.3 a scheme of the principal power, temperature and pressure data of the KALIMER-600 is shown. Figure 1.3: KALIMER-600 Steady state heat balance 7

9 Table 1.4 summarizes the key design parameters of KALIMER-600. Table 1.4: Kalimer-600 Key Design Parameters 8

10 1.3 TECHNICAL PROBLEMS Since numerous sodium cooled reactors have been built and operated in different countries (Japan, France, Germany, United Kingdom, Russia and USA) in the past decades, the approach for the development of a Sodium Fast Reactor is to rely on technologies already used. For this reason a significant fraction of the R&D funds at disposal for the SFR are related to performance rather than viability of the project. In spite of this advantage for the early readiness of the system like a mature technology to be introduced on the market, some technology challenges to resolve viability issues remain and can be summarized as follows: Advanced fuel and fuel cycles o The completion of the fuel database o The fabrication of ceramic pellets containing minor actinides and trace amounts of fission products o The scale-up of the uranium crystallization step o The pyroprocess: viability issues include the development and demonstration of engineering-scale recovery of transuranics, and the demonstration of large throughout operations Components design o the capital cost reduction to levels competitive with those of other nuclear systems and power sources o development of advanced in-service inspection and repair technologies (in presence of sodium) Design and safety o the verification of the predictability and the effectiveness of the mechanisms that contribute to inherently safe response to design basis transients and anticipated transients without scram o the certainty that bounding events considered in licensing can be sustained without loss of coolability of fuel The main problem related to sodium is that the sodium/water reaction is highly exothermic and, because energy conversion systems typically use water-steam Rankine cycles, the sodium-water/steam interface must be maintained sufficiently separated from the primary sodium coolant, so that the core is not affected in case 9

11 of reaction. This has resulted in the addition of an intermediate sodium loop, with certain negative impact in the economics of the plant. It could be possible to avoid the presence of the intermediate heat exchange loop, and at the same time simplifying the system and reducing the capital costs of the plant modifying the cycle for the power generation. It means the use of a supercritical CO 2 Brayton cycle instead of a Water/Steam Rankine Cycle. This idea is not only applicable to Sodium Fast Reactors, but in this case the modification of the general layout of the plant will be quite effective. Because sodium systems must be maintained isolated form air and continuously sealed, design of the plant and specific components for high reliability and reduced or simplified maintenance is of particular importance. The long term objective is extended operation with low maintenance requirements. This becomes even more relevant for plant designs for long life (60 years or more). For this reason materials must be qualified to operate reliably at operating temperatures for extended periods of time. This may require extending the database for materials currently in use. Performing inspection and repair of sodium systems presents the challenges of overcoming the coolant opacity, its relatively high temperature and the need to maintain inert atmosphere. The technical problems and what is left to be done for the SFR could be resumed essentially in safety goals related to increase the operational safety and reliability of the system like: minimization of the frequency and severity of severe accidents resulting in avoiding offsite emergency response design simplicity and the coolant/structural advanced in-service inspection and maintenance the use of inherent and passive safety will allow the accommodation of operational transients and severe accidents and will also preclude any significant damage the minimization of sodium leaks: 10

12 There is extensive experience with sodium leak events, but most of them have been very minor. There have been no core damage events induced by sodium leak. It is important to reduce the possibility of the sodium leak as much as possible by taking the experience of the sodium leak events into account from the view point of safety as well as the plant performance. 11

13 1.4 ECONOMIC ASPECTS Economics is a key issue for SFR reactors. Commodities in a SFR are usually larger than for a light water reactor of a similar power and also the electricity generation costs. A major effort must be undertaken in advanced SFR design to make the plants competitive. These considerations are still valid for sodium cooled reactors since an intermediate heat exchange loop has to be interposed between the primary and the secondary loop for safety reasons (highly exothermic sodium/water reaction is possible). If a supercritical CO 2 Brayton cycle is used, the intermediate cooling loop could be omitted and the SFR could become more economically advantageous. Further research is needed to confirm the viability of this approach. Development of advanced materials could permit a more compact design. This will represent a big step towards lower capital costs. In the same way, development of advanced components for high reliability and long live will be a significant improvement. To achieve this goal a very important role should be played by advanced in-service inspection and repair technologies and by the development of advanced components. This could lead to ensuring a long live for the plant (lower costs) and lower operational and maintenance costs as well. Key issues for SFR systems are achieving a capital cost reduction of the power plant by innovative engineering design, the use of advanced maintenance technologies and advanced fuel cycle in order to lower operating and fuel/disposal costs. These cost reductions should allow the SFRs to be competitive in capital costs and in the costs per kilowatt hour of producing electricity (busbar costs) in comparison with those of advanced LWR systems. Further, we should not forget that in future advanced energy conversion systems will be on hand and they will potentially increase the thermal efficiency and lower the capital costs. Sodium cooled fast reactors have been also successfully operated for the sea water desalination. The BN-350 was a sodium-cooled fast reactor nuclear power plant located at Aktau (formally known as Shevchenko from ), Kazakhstan, situated on the shore of the Caspian Sea. The town was settled first to be the camp of the workers of oil industry from another parts of the country emerged on the 12

14 Mangyshlak peninsula, in one of the most lifeless deserts in Kazakhstan. In fact, in addition to providing power for the city (150 MWe) and for producing plutonium, the BN-350 was also used for desalination to supply fresh water ( m³ fresh water/day) to the city. The presence of the reactor permitted the development of the population since a source of drinking water is not present near the city. Nowadays, the population of the city reaches the value of more or less inhabitants. By the long-term plan of the Government of Kazakhstan a new nuclear power station is going to be built near the acutal, as the current type of supplying water and energy will not be enough in the future, since the new district called Aktau - City is going to be built, doubling the current area of the city. 13

15 1.5 ENVIRONMENTAL AND SOCIAL ASPECTS The category of the Fast Reactors is far away to reach a level of normal social acceptance. Many environmental associations consider Fast Reactors as unsafe, uneconomical, and unable to resolve the problems of nuclear power. Many of them are using partial and incorrect source to justify their positions, even if in some cases they are addressing their attention to the open question to be solved by future R&D activities in the field to achieve better targets in the economic and safety of these kind reactors. A SFR system should rely on enhancing intrinsic non-proliferation characteristics of the fuel cycle. It means the use of advanced recycling technologies that avoid essentially the separation of plutonium, operating with low decontamination factors of the recycled products. It is also important the minimization of nuclear material transportation, and facilitating the application of effective international safeguards over the entire lifecycle of materials. The need for long-term safeguards for the repository is eliminated by avoiding the disposal of actinides. The primary mission is the management of nuclear waste, in particular management of plutonium and other actinides. SFR allows the utilization of almost all the energy in the natural uranium, versus the 1% utilization in thermal spectrum systems. The long term toxicity of the waste sent to the repository is enormously reduced by eliminating essentially all actinides, since the SFR system allows for full recycle of actinides and for adaptation of the core to actinide management needs (initially operating in a net burning mode to consume potential excess TRU material (Trans-Uranic Actinides), and switching in the future to a net fissile creation mode to enhance sustainability). In case of operation without fuel rod failures, after-irradiation cooling of sodium for years would be enough to use sodium again in any way or to return it in the environment The case of the attack to the French Fast Reactor Superphénix is emblematic to explain the position of the anti nuclear activists. During the 1982 a rocket attack was launched against the unfinished plant by an "eco-pacifist group". Five rockets were launched using a Russian rocket launcher. The incomplete containment building was damaged by two of the rockets, which narrowly missed the reactor's empty core. 14

16 Also the case of the German SNR-300 breeder reactor is symbolic. In 1985 the reactor was ready to receive nuclear materials and to be operated but it was taken in partial operation. The sodium coolant was already running through the coolant loop and had to be kept hot using electric heating elements so it did not solidify. The cause of termination of the SNR-300 project is attributed to the Authorities of the State of North Rhine-Westphalia. The final cancellation of the program arrived during the

17 1.6 BIBLIOGRAPHY Na Cooled Reactors : Technological Challenges and New Designs, Jordi Roglans Ribas, Yutaka Sagayama, The 2004 Frédéric JOLIOT & Otto HAHN Summer School AUGUST 25 SEPTEMBER 3, 2004 CADARACHE, France Design Concept of KALIMER-600 Dohee HAHN, Yeong-Il KIM, Seong-O KIM, Jae- Han LEE and Yong-Bum LEE Proceedings of GLOBAL 2005 Tsukuba, Japan, Oct 9-13, 2005 Paper No. 437 IAEA Fast Reactor Database - INL FY2005 Report, Appendix 5.0 Sodium-Cooled Fast Reactors, Idaho National Laboratory, USA, 2005 Comparison of sodium and lead-cooled fast reactors regarding severe safety and economical issues, Kamil Tucek, Johan Carlsson, Hartmut Wider, 13th International Conference on Nuclear Engineering, Beijing, China, May 16-20, 2005 ICONE Liquid Metal Cooled Reactors: Experience in Design and Operation, IAEA, WIEN, 2007, IAEA-TECDOC