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Latest liquid lithium target design during the key element technology phase in the international fusion materials irradiation facility (IFMIF)
Fusion Engineering and Design 66 /68 (2003) 193 /198 www.elsevier.com/locate/fusengdes
Latest liquid lithium target design during the key element technology phase in the international fusion materials irradiation facility (IFMIF) Hiroo Nakamura a,*, B. Riccardi b, K. Ara c, L. Burgazzi d, S. Cevolani d, G. Dell’Orco e, C. Fazio e, D. Giusti d, H. Horiike f, M. Ida a, H. Ise g, H. Kakui h, N. Loginov i, H. Matsui j, T. Muroga k, Hideo Nakamura a, K. Shimizu l, H. Takeuchi a, S. Tanaka m a
Office of Fusion Materials Research Promotion, Tokai Research Establishment, Japan Atomic Energy Research Institute, 2-4 Shirakata-Shirane, Tokai-mura, Ibaraki-ken 319-1195, Japan b Associazione EURATOM ENEA sulla Fusione, I-00044 Frascati, Rome, Italy c Japan Nuclear Cycle Development Institute, Ibaraki 311-1394, Japan d ENEA CR ‘E.Clemental’, I-40129 Bologna, Italy e ENEA CR Brasimone, I-40035 Camugnano (Bo), Italy f Osaka University, Osaka 565-0871, Japan g Kawasaki Heavy Industry, Tokyo 136-8588, Japan h Ishikawajima-Harima Heavy Industry Co., Ltd., Tokyo 100-8182, Japan i Institute for Physics and Power Engineering, Obninsk 249020, Russia j IMR, Tohoku University, Miyagi 908-0812, Japan k National Institute for Fusion Science, Gifu 509-5292, Japan l Mitsubishi Heavy Industry, Hyogo 652-8585, Japan m University of Tokyo, Tokyo 113-8656, Japan
Abstract International Fusion Materials Irradiation Facility (IFMIF), being jointly developed by EU, JA, RF and US, is a deuteron /lithium (Li) stripping reaction neutron source for fusion materials testing. In 2002, a 3 year Key Element technology Phase (KEP) to reduce the key technology risk factors was completed. A liquid Li target has been designed to produce intense high energy neutrons (2 MW/m2) up to 50 dpa/year by 10 MW of deuterium beam deposition which corresponds to an ultra high heat load of 1 GW/m2. This paper describes the latest design of the liquid Li target system reflecting the KEP results and future prospects. # 2003 Elsevier B.V. All rights reserved. Keywords: Lithium; Irradiation; Key element technology phase; IFMIF
* Corresponding author. Tel.: /81-29-284-3791; fax: /81-29-282-5551. E-mail address: [email protected] (H. Nakamura). 0920-3796/03/$ - see front matter # 2003 Elsevier B.V. All rights reserved. doi:10.1016/S0920-3796(03)00208-4
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1. Introduction The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D/Li neutron source to produce high energy neutron spectrum (peaked at 14 /16 MeV), high neutron flux (2 MW/m2) reaching /20 dpa/year in a sufficient irradiation volume (500 cm3) for testing the candidate materials used in DEMO [1]. To realize such a condition, two 40 MeV deuteron beams, each with 125 mA beam current, impinge a high speed liquid Li flow target. Following the Conceptual Design Activity (CDA) [2,3] and cost reduction study [4], a 3 year Key Element technology Phase (KEP) was initiated in 2000 to reduce the key technology risk factors in the IFMIF design. This paper describes the latest design of the liquid Li target system reflecting the KEP results [5 /7]. Future prospects are also summarized.
2. Requirements of IFMIF target system Design requirements of the Li target system are: heat removal of 10 MW from the deuterium beams, production of a stable Li jet for production of intense neutron irradiation damage more than 20 dpa/year in the high flux test module, control of the impurity level (T, 7Be, C, O, N) below permissible levels, ensuring safety with respect to Li hazard and tritium release from the Li loop and achievement of system availability of more than 95% during the lifetime of the plant [8]. Major specifications of the IFMIF Li target are summarized in Table 1. The heat flux of 1 GW/m2 is the averaged value in the beam footprint on the free liquid Li flow. To handle such an ultra high heat load without Li boiling and significant evaporation, flowing liquid Li target is necessary. To provide a stable intense neutron source, amplitude waves of the Li free surface must be less than 1 mm. The target assembly and/or back wall must be replaceable by a remote handling system with a setting accuracy of 0.5 mm considering the small gap (2 mm) between the back wall and the test cell assembly. To ensure safe operation of the Li target system, diagnostics on Li temperature, distortion
Table 1 Major specification of IFMIF lithium target Items
Values
Beam power/average heat flux Beam deposition area on Li jet Width/thickness of Li jet Velocity of Li jet Flow rate of Li Inlet temperature of Li Permissible wave amplitude Pressure in front of free surface Permissible corrosion thickness Permissible impurity contents Tritium generation rate Materials (back wall)
10 MW/1 GW/m2
(Other components) Replacement (backwall) (Other components) Accuracy of installation
200 mm (width) /50 mm (height) 260/25 mm 15 (range 10 /20) m/s 130 l/s 250 8C 1 mm (Li free surface) 10 3 Pa 500 mm/40 years (piping) 10 wppm (C, N, O; each) 7 g/year Reduced-activation ferritic/martensitic steel 304 stainless steel 11 month/fpy No replacement for 20 years 0.5 mm (target assembly)
of the target assembly and back wall, Li leakage and tritium release is required.
3. Latest design 3.1. Target assembly and Li loop The Li target system consists of two main components: the target assembly and the Li loop. Block diagram of the target system is shown in Fig. 1. The Li loop circulates the Li to and from the target assembly through the Li purification system and heat exchange system by an electromagnetic pump. The Li purification system consists of a cold trap and two hot traps. It is able to maintain tritium, 7Be and other impurities under permissible values. Fig. 2 shows a cross section of the Li target assembly. The target assembly is stainless steel alloy except for the back wall made of Reduced-Activation Ferritic/Martensitic (RAFM) steel. The flow straightener is aimed to
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To evaluate the transient thermal behavior of the Li loop in case of deuterium beam trip or startup, transient analysis has been performed. The result shows that without any control of the Li temperature by a heat exchanger system, Li temperature at the exit of the Li cooler the temperature becomes lower than the solidification temperature of Li (180 8C) after 380 s. 3.2. Lithium purification system
Fig. 1. Block diagram of the IFMIF Li target system.
change turbulent flow to laminar flow. At an exit of the straightner, the double-reducer nozzle is placed. This nozzle has a large contraction ratio of 10; flow separation was avoided by employing in the nozzle design Shima’s model based on the potential flow theory. One of the design issues in the target assembly is the thermal-hydrodynamic behavior of the Li flow in the double-reducer nozzle. Multi-dimensional analysis has been performed using FLOW-3D code [9]. Typical result shows that a maximum temperature in the Li flow is about 400 8C. To avoid boiling of the Li, the pressure on the liquid is increased by a centrifugal force; therefore, a curvature of 25 cm is adopted in the back wall. Since the boiling temperature increases up to 1090 8C under an acceleration of 160 G due to the curved back wall, enough margin is present to avoid a Li boiling in the Li flow. Characteristics of the Li jet flow has been simulated in the water jet experiment; at 0.01 MPa stable water jet was observed up to 20 m/s [10]. In order to validate the double-reducer nozzle, Li jet flow experiment has been started [11]. In the target assembly, the back wall must withstand the most severe neutron irradiation condition (50 dpa/year), therefore, the back wall needs to be exchanged remotely every 11 month.
Major impurities in the Li loop are protium, deuterium, tritium, 7Be, and other species (C, N, O). T and 7Be (a half-life of 53 days) are the dominant hazardous source term in the event of a radiological release and are produced by direct reactions of the deuterium beam with the Li. Deuterium from the beam is also contained in the Li flow. The total estimated production rates of H, D, and T for a full power operation are about 5, 160, 7 g/year, respectively. Li purification system consists of a cold trap, a V /10% Ti (and/or Cr)-hot trap [12], an Yttrium hot trap and on-line/ off-line monitors. Instead of Titanium, V /10% Ti (and/or Cr) is considered to be the candidate material of the hot trap to control the nitrogen content at 550 8C up to 600 8C. Hydrogen isotopes (H, D, T) are removed by the Yttrium hot trap at 285 8C [13]. As an option, a cold trap with protium sparging, so-called ‘‘swamping method’’ is also considered. Without removal of 7Be, saturated activation level becomes 140 kCi. Although 7Be is removed by the cold trap at 200 8C, some material will remain inside the loop and, therefore, a remote handling system need to be considered for maintenance. Impurities of C, O and N are generated from the loop materials. The cold trap will remove C and the V /10% Ti hot trap at 550 8C will remove O and N. In addition to the erosion/ corrosion control, the reduction of the nitrogen and oxygen impurities is needed to avoid deterioration of the Yttrium hot trap. The design value of the permissible concentration of these species is now selected to be 10 wppm for each impurity. It will be updated on the basis of the KEP results. The impurity monitoring system consists of on-line and off-line systems. These systems will include hydrogen membrane diffusion meter and a resis-
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Fig. 2. Cross section of target assembly.
tivity monitor. Since the on-line monitors need an extensive development, their need should be assessed. 3.3. Remote handling system The back wall made of RAFM steel is operating under severe condition of neutron irradiation damage (about 50 dpa/year), therefore, the back wall is designed for replacement every 11 months. Two design options of the back wall replacement are under investigation. The first option is based on the removal of the overall target assembly with the back wall to the hot cell area for replacement of the back wall itself (see Fig. 3). YAG laser device will be used for cutting the lip seals of the
flanges. The remote handling system to exchange the target assembly is unified into the Universal Robot System (URS) of the IFMIF Test Facilities. The study of the procedure of the target assembly transfer system to the hot cell is in progress. The second option is to replace the back wall of the target assembly by a remote handling device [6]. In this case, ‘‘bayonet’’ type mechanical attachment of the back wall to the target assembly is envisaged. R&D activities on remote handling are under way. 3.4. Diagnostics In order to control the Li target system during operation, conceptual design of the monitors of
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tion of the target system over a prolonged operation time (more than 10 000 h) will be performed. These issues include (1) validation of the Li flow stability with an IFMIF scale nozzle and a concave back wall. Transient behavior of the Li flow simulating IFMIF operation will be investigated, (2) validation of the Li loop layout to verify absence of the cavitation and operation under vacuum condition of 10 3 Pa, (3) validation of compatibility of the Li loop materials under a low impurity level of 10 wppm for each impurity species, (4) validation of the Li purification system and of diagnostics and (5) demonstration of safety operation of the IFMIF Li loop.
4. Summary Fig. 3. Remote handling of the IFMIF target (option-1).
the Li temperature, Li flow velocity, Li thickness and displacement in the target assembly and the back wall has been carried out. Temperature of the Li free surface is measured by infra-red (IR) cameras, thermo-couples and ultra sonic sensors. The IR camera is used to monitor the location of the deuterium beam footprint and surface temperature of the Li. The thermocouples are used to measure the Li temperature of the flow and to calibrate the IR camera. The Li flow velocity and thickness are measured by the ultra sonic sensors. Ultra sonic sensors are also used to measure the Li temperature inside the flow. The displacement of the target assembly and back wall are measured by a laser diagnostics with a spatial resolution of 0.1 mm, located at 15 m from the Li target. 3.5. Activities in KEP and future prospects In EU, Japan and Russian Federation, the KEP tasks on stability of Li flow, damage/corrosion by Li flow, Li purification, Li vaporization, safety analysis, loop integrity and remote handling have are in progress. The KEP activities have been accomplished on December 2002 [7]. Following the KEP activities, in the next phase, the ‘‘Engineering Validation and Engineering Design Activities (EVEDA)’’, the engineering valida-
Design of the IFMIF target system has been performed to handle an ultra high heat load of 1 GW/m2 due to the injection of 10 MW deuterium beams. A high speed Li flow of 20 m/s and a concave back wall are applied to the target design. KEP activities on the water jet and Li loop facilities have been reported. After the KEP, 5 years of EVEDA will validate the critical components mainly with respect to reliability and availability.
Acknowledgements Appreciation is given to all members of the IFMIF international team on their contributions to this work. The IFMIF executive subcommittee has been highly impressed with dedication and enthusiasm of the IFMIF team. Also, appreciation is given to for the continued interest of the IEAFPCC and IEA Executive Committee on Fusion Material.
References [1] A. Mo¨slang, U. Fischer, V. Heinzel, P. Vladimirov, R. Ferdinand, H. Klein, et al., Proc. 19th Fusion Energy Conference, France, October 2002, IAEA-CN-94/FT1-2.
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[2] IFMIF-CDA Team, M. Martone (Ed.), IFMIF Conceptual Design Activity, Final Report, ENEA Frascati Report, RT/ERG/FUS/96/17, 1996. [3] IFMIF Team, A. Moeslang (Ed.), IFMIF Conceptual Design Evaluation Report, Forschungszentrum Karlsruhe report, FZKA 6199, January 1999. [4] H. Takeuchi, M. Sugimoto, H. Nakamura, T. Yutani, M. Ida, S. Jitsukawa, T. Kando, et al., Proceedings of the 18th Fusion Energy Conference, Italy, October 2000, IAEACN-77/FTP2/03. [5] Annual Report of NIFS Collaboration on Key Element Technology Development for Fusion Materials Irradiation Test facility, January 2001, National Institute for Fusion Science, in Japanese. [6] B. Riccardi, M. Martone, C. Antonucci, L. Burgazzi, S. Cevolani, D. Guisti, G. Dell’Orco, et al., 22nd Symposium on Fusion Technology, Helsinki, 9 /13 September 2002, Fusion Eng. Des., PII: S0920-3796(03)00209-6. [7] IFMIF team, IFMIF KEP Report, JAERI Laboratory report, JAERI-Tech 2003 /005, March 2003.
[8] H. Nakamura, M. Ida, M. Sugimoto, H. Takeuchi, T. Yutani, International IFMIF Team, Fusion Eng. Des. 58 / 59 (2001) 919 /923. [9] H. Nakamura, K. Itih, Y. Kukita, M. Ida, Y. Kaoto, H. Maekawa, H. Katsuta, J. Nucl. Mater. 258 /263 (1998) 440 /445. [10] M. Ida, H. Horiike, M. Akiba, K. Ezato, T. Iida, S. Inoue, S. Miyamoto, et al., Proceedings of the Tenth International Conference on Fusion Reactor Materials, BadenBaden, 14 /19 October 2001, J. Nucl. Mater., 307-311 (2002) 1686 /1690. [11] H. Horiike, M. Ida, T. Iida, S. Inoue, S. Miyamoto, T. Muroga, Hideo Nakamura, et al., 22nd Symposium on Fusion Technology, Helsinki, 9 /13 September 2002, Fusion Eng. Des., PII: S0920-3796(03)00205-9. [12] T. Sakurai, T. Yoneoka, S. Tonaka, A. Suzuki, T. Muroga, Proceedings of the Sixth International Symposium on Fusion Nuclear Technology, San Diego, April 2002, Fusion Eng. Des., 61-62 (2002) 763 /768. [13] H. Nakamura, M. Ida, M. Sugimoto, T. Yutani, H. Takeuchi, Fusion Sci. Technol. 41 (2002) 845.
Latest liquid lithium target design during the key element technology phase in the international fusion materials irradiation facility (IFMIF) Hiroo Nakamura a,*, B. Riccardi b, K. Ara c, L. Burgazzi d, S. Cevolani d, G. Dell’Orco e, C. Fazio e, D. Giusti d, H. Horiike f, M. Ida a, H. Ise g, H. Kakui h, N. Loginov i, H. Matsui j, T. Muroga k, Hideo Nakamura a, K. Shimizu l, H. Takeuchi a, S. Tanaka m a
Office of Fusion Materials Research Promotion, Tokai Research Establishment, Japan Atomic Energy Research Institute, 2-4 Shirakata-Shirane, Tokai-mura, Ibaraki-ken 319-1195, Japan b Associazione EURATOM ENEA sulla Fusione, I-00044 Frascati, Rome, Italy c Japan Nuclear Cycle Development Institute, Ibaraki 311-1394, Japan d ENEA CR ‘E.Clemental’, I-40129 Bologna, Italy e ENEA CR Brasimone, I-40035 Camugnano (Bo), Italy f Osaka University, Osaka 565-0871, Japan g Kawasaki Heavy Industry, Tokyo 136-8588, Japan h Ishikawajima-Harima Heavy Industry Co., Ltd., Tokyo 100-8182, Japan i Institute for Physics and Power Engineering, Obninsk 249020, Russia j IMR, Tohoku University, Miyagi 908-0812, Japan k National Institute for Fusion Science, Gifu 509-5292, Japan l Mitsubishi Heavy Industry, Hyogo 652-8585, Japan m University of Tokyo, Tokyo 113-8656, Japan
Abstract International Fusion Materials Irradiation Facility (IFMIF), being jointly developed by EU, JA, RF and US, is a deuteron /lithium (Li) stripping reaction neutron source for fusion materials testing. In 2002, a 3 year Key Element technology Phase (KEP) to reduce the key technology risk factors was completed. A liquid Li target has been designed to produce intense high energy neutrons (2 MW/m2) up to 50 dpa/year by 10 MW of deuterium beam deposition which corresponds to an ultra high heat load of 1 GW/m2. This paper describes the latest design of the liquid Li target system reflecting the KEP results and future prospects. # 2003 Elsevier B.V. All rights reserved. Keywords: Lithium; Irradiation; Key element technology phase; IFMIF
* Corresponding author. Tel.: /81-29-284-3791; fax: /81-29-282-5551. E-mail address: [email protected] (H. Nakamura). 0920-3796/03/$ - see front matter # 2003 Elsevier B.V. All rights reserved. doi:10.1016/S0920-3796(03)00208-4
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1. Introduction The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D/Li neutron source to produce high energy neutron spectrum (peaked at 14 /16 MeV), high neutron flux (2 MW/m2) reaching /20 dpa/year in a sufficient irradiation volume (500 cm3) for testing the candidate materials used in DEMO [1]. To realize such a condition, two 40 MeV deuteron beams, each with 125 mA beam current, impinge a high speed liquid Li flow target. Following the Conceptual Design Activity (CDA) [2,3] and cost reduction study [4], a 3 year Key Element technology Phase (KEP) was initiated in 2000 to reduce the key technology risk factors in the IFMIF design. This paper describes the latest design of the liquid Li target system reflecting the KEP results [5 /7]. Future prospects are also summarized.
2. Requirements of IFMIF target system Design requirements of the Li target system are: heat removal of 10 MW from the deuterium beams, production of a stable Li jet for production of intense neutron irradiation damage more than 20 dpa/year in the high flux test module, control of the impurity level (T, 7Be, C, O, N) below permissible levels, ensuring safety with respect to Li hazard and tritium release from the Li loop and achievement of system availability of more than 95% during the lifetime of the plant [8]. Major specifications of the IFMIF Li target are summarized in Table 1. The heat flux of 1 GW/m2 is the averaged value in the beam footprint on the free liquid Li flow. To handle such an ultra high heat load without Li boiling and significant evaporation, flowing liquid Li target is necessary. To provide a stable intense neutron source, amplitude waves of the Li free surface must be less than 1 mm. The target assembly and/or back wall must be replaceable by a remote handling system with a setting accuracy of 0.5 mm considering the small gap (2 mm) between the back wall and the test cell assembly. To ensure safe operation of the Li target system, diagnostics on Li temperature, distortion
Table 1 Major specification of IFMIF lithium target Items
Values
Beam power/average heat flux Beam deposition area on Li jet Width/thickness of Li jet Velocity of Li jet Flow rate of Li Inlet temperature of Li Permissible wave amplitude Pressure in front of free surface Permissible corrosion thickness Permissible impurity contents Tritium generation rate Materials (back wall)
10 MW/1 GW/m2
(Other components) Replacement (backwall) (Other components) Accuracy of installation
200 mm (width) /50 mm (height) 260/25 mm 15 (range 10 /20) m/s 130 l/s 250 8C 1 mm (Li free surface) 10 3 Pa 500 mm/40 years (piping) 10 wppm (C, N, O; each) 7 g/year Reduced-activation ferritic/martensitic steel 304 stainless steel 11 month/fpy No replacement for 20 years 0.5 mm (target assembly)
of the target assembly and back wall, Li leakage and tritium release is required.
3. Latest design 3.1. Target assembly and Li loop The Li target system consists of two main components: the target assembly and the Li loop. Block diagram of the target system is shown in Fig. 1. The Li loop circulates the Li to and from the target assembly through the Li purification system and heat exchange system by an electromagnetic pump. The Li purification system consists of a cold trap and two hot traps. It is able to maintain tritium, 7Be and other impurities under permissible values. Fig. 2 shows a cross section of the Li target assembly. The target assembly is stainless steel alloy except for the back wall made of Reduced-Activation Ferritic/Martensitic (RAFM) steel. The flow straightener is aimed to
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To evaluate the transient thermal behavior of the Li loop in case of deuterium beam trip or startup, transient analysis has been performed. The result shows that without any control of the Li temperature by a heat exchanger system, Li temperature at the exit of the Li cooler the temperature becomes lower than the solidification temperature of Li (180 8C) after 380 s. 3.2. Lithium purification system
Fig. 1. Block diagram of the IFMIF Li target system.
change turbulent flow to laminar flow. At an exit of the straightner, the double-reducer nozzle is placed. This nozzle has a large contraction ratio of 10; flow separation was avoided by employing in the nozzle design Shima’s model based on the potential flow theory. One of the design issues in the target assembly is the thermal-hydrodynamic behavior of the Li flow in the double-reducer nozzle. Multi-dimensional analysis has been performed using FLOW-3D code [9]. Typical result shows that a maximum temperature in the Li flow is about 400 8C. To avoid boiling of the Li, the pressure on the liquid is increased by a centrifugal force; therefore, a curvature of 25 cm is adopted in the back wall. Since the boiling temperature increases up to 1090 8C under an acceleration of 160 G due to the curved back wall, enough margin is present to avoid a Li boiling in the Li flow. Characteristics of the Li jet flow has been simulated in the water jet experiment; at 0.01 MPa stable water jet was observed up to 20 m/s [10]. In order to validate the double-reducer nozzle, Li jet flow experiment has been started [11]. In the target assembly, the back wall must withstand the most severe neutron irradiation condition (50 dpa/year), therefore, the back wall needs to be exchanged remotely every 11 month.
Major impurities in the Li loop are protium, deuterium, tritium, 7Be, and other species (C, N, O). T and 7Be (a half-life of 53 days) are the dominant hazardous source term in the event of a radiological release and are produced by direct reactions of the deuterium beam with the Li. Deuterium from the beam is also contained in the Li flow. The total estimated production rates of H, D, and T for a full power operation are about 5, 160, 7 g/year, respectively. Li purification system consists of a cold trap, a V /10% Ti (and/or Cr)-hot trap [12], an Yttrium hot trap and on-line/ off-line monitors. Instead of Titanium, V /10% Ti (and/or Cr) is considered to be the candidate material of the hot trap to control the nitrogen content at 550 8C up to 600 8C. Hydrogen isotopes (H, D, T) are removed by the Yttrium hot trap at 285 8C [13]. As an option, a cold trap with protium sparging, so-called ‘‘swamping method’’ is also considered. Without removal of 7Be, saturated activation level becomes 140 kCi. Although 7Be is removed by the cold trap at 200 8C, some material will remain inside the loop and, therefore, a remote handling system need to be considered for maintenance. Impurities of C, O and N are generated from the loop materials. The cold trap will remove C and the V /10% Ti hot trap at 550 8C will remove O and N. In addition to the erosion/ corrosion control, the reduction of the nitrogen and oxygen impurities is needed to avoid deterioration of the Yttrium hot trap. The design value of the permissible concentration of these species is now selected to be 10 wppm for each impurity. It will be updated on the basis of the KEP results. The impurity monitoring system consists of on-line and off-line systems. These systems will include hydrogen membrane diffusion meter and a resis-
196
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Fig. 2. Cross section of target assembly.
tivity monitor. Since the on-line monitors need an extensive development, their need should be assessed. 3.3. Remote handling system The back wall made of RAFM steel is operating under severe condition of neutron irradiation damage (about 50 dpa/year), therefore, the back wall is designed for replacement every 11 months. Two design options of the back wall replacement are under investigation. The first option is based on the removal of the overall target assembly with the back wall to the hot cell area for replacement of the back wall itself (see Fig. 3). YAG laser device will be used for cutting the lip seals of the
flanges. The remote handling system to exchange the target assembly is unified into the Universal Robot System (URS) of the IFMIF Test Facilities. The study of the procedure of the target assembly transfer system to the hot cell is in progress. The second option is to replace the back wall of the target assembly by a remote handling device [6]. In this case, ‘‘bayonet’’ type mechanical attachment of the back wall to the target assembly is envisaged. R&D activities on remote handling are under way. 3.4. Diagnostics In order to control the Li target system during operation, conceptual design of the monitors of
H. Nakamura et al. / Fusion Engineering and Design 66 /68 (2003) 193 /198
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tion of the target system over a prolonged operation time (more than 10 000 h) will be performed. These issues include (1) validation of the Li flow stability with an IFMIF scale nozzle and a concave back wall. Transient behavior of the Li flow simulating IFMIF operation will be investigated, (2) validation of the Li loop layout to verify absence of the cavitation and operation under vacuum condition of 10 3 Pa, (3) validation of compatibility of the Li loop materials under a low impurity level of 10 wppm for each impurity species, (4) validation of the Li purification system and of diagnostics and (5) demonstration of safety operation of the IFMIF Li loop.
4. Summary Fig. 3. Remote handling of the IFMIF target (option-1).
the Li temperature, Li flow velocity, Li thickness and displacement in the target assembly and the back wall has been carried out. Temperature of the Li free surface is measured by infra-red (IR) cameras, thermo-couples and ultra sonic sensors. The IR camera is used to monitor the location of the deuterium beam footprint and surface temperature of the Li. The thermocouples are used to measure the Li temperature of the flow and to calibrate the IR camera. The Li flow velocity and thickness are measured by the ultra sonic sensors. Ultra sonic sensors are also used to measure the Li temperature inside the flow. The displacement of the target assembly and back wall are measured by a laser diagnostics with a spatial resolution of 0.1 mm, located at 15 m from the Li target. 3.5. Activities in KEP and future prospects In EU, Japan and Russian Federation, the KEP tasks on stability of Li flow, damage/corrosion by Li flow, Li purification, Li vaporization, safety analysis, loop integrity and remote handling have are in progress. The KEP activities have been accomplished on December 2002 [7]. Following the KEP activities, in the next phase, the ‘‘Engineering Validation and Engineering Design Activities (EVEDA)’’, the engineering valida-
Design of the IFMIF target system has been performed to handle an ultra high heat load of 1 GW/m2 due to the injection of 10 MW deuterium beams. A high speed Li flow of 20 m/s and a concave back wall are applied to the target design. KEP activities on the water jet and Li loop facilities have been reported. After the KEP, 5 years of EVEDA will validate the critical components mainly with respect to reliability and availability.
Acknowledgements Appreciation is given to all members of the IFMIF international team on their contributions to this work. The IFMIF executive subcommittee has been highly impressed with dedication and enthusiasm of the IFMIF team. Also, appreciation is given to for the continued interest of the IEAFPCC and IEA Executive Committee on Fusion Material.
References [1] A. Mo¨slang, U. Fischer, V. Heinzel, P. Vladimirov, R. Ferdinand, H. Klein, et al., Proc. 19th Fusion Energy Conference, France, October 2002, IAEA-CN-94/FT1-2.
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[2] IFMIF-CDA Team, M. Martone (Ed.), IFMIF Conceptual Design Activity, Final Report, ENEA Frascati Report, RT/ERG/FUS/96/17, 1996. [3] IFMIF Team, A. Moeslang (Ed.), IFMIF Conceptual Design Evaluation Report, Forschungszentrum Karlsruhe report, FZKA 6199, January 1999. [4] H. Takeuchi, M. Sugimoto, H. Nakamura, T. Yutani, M. Ida, S. Jitsukawa, T. Kando, et al., Proceedings of the 18th Fusion Energy Conference, Italy, October 2000, IAEACN-77/FTP2/03. [5] Annual Report of NIFS Collaboration on Key Element Technology Development for Fusion Materials Irradiation Test facility, January 2001, National Institute for Fusion Science, in Japanese. [6] B. Riccardi, M. Martone, C. Antonucci, L. Burgazzi, S. Cevolani, D. Guisti, G. Dell’Orco, et al., 22nd Symposium on Fusion Technology, Helsinki, 9 /13 September 2002, Fusion Eng. Des., PII: S0920-3796(03)00209-6. [7] IFMIF team, IFMIF KEP Report, JAERI Laboratory report, JAERI-Tech 2003 /005, March 2003.
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