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JAEA Reports

Assessment report of research and development activities in FY2014; Activity "Fusion Research and Development" (In-advance evaluation)

Fusion Research and Development Directorate

JAEA-Evaluation 2016-002, 40 Pages, 2016/03

JAEA-Evaluation-2016-002.pdf:2.66MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") asked the assessment committee, "Evaluation Committee of Research and Development Activities for Fusion" (hereinafter referred to as "Committee") for in-advance evaluation of "Research and Development of the technical system for extraction of fusion energy," in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and "Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to the JAEA's request, the Committee assessed the research program of the Fusion Research and Development Directorate (hereinafter referred to as "FRDD") during the period from April 2015 to March 2022. The Committee evaluated the management and research activities of the FRDD based on the explanatory documents prepared by the FRDD, the oral presentations with questions-and-answers by the Director General and the Deputy Director Generals.

JAEA Reports

Assessment report of research and development activities in FY2014; Activity "Fusion Research and Development" (Result evaluation)

Fusion Research and Development Directorate

JAEA-Evaluation 2016-001, 128 Pages, 2016/03

JAEA-Evaluation-2016-001.pdf:33.25MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") asked the assessment committee, "Evaluation Committee of Research and Development Activities for Fusion" (hereinafter referred to as "Committee") for result evaluation of "Research and Development of the Technical System for Extraction of Fusion Energy," in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology " and "Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to the JAEA's request, the Committee assessed the research program of the Fusion Research and Development Directorate (hereinafter referred to as "FRDD") during the period from April 2010 to November 2014. The Committee evaluated the management and research activities of the FRDD based on the explanatory documents prepared by the FRDD, the oral presentations with questions-and-answers by the Director General and the Deputy Director Generals.

Journal Articles

Non-destructive examination of jacket sections for ITER central solenoid conductors

Takahashi, Yoshikazu; Suwa, Tomone; Nabara, Yoshihiro; Ozeki, Hidemasa; Hemmi, Tsutomu; Nunoya, Yoshihiko; Isono, Takaaki; Matsui, Kunihiro; Kawano, Katsumi; Oshikiri, Masayuki; et al.

IEEE Transactions on Applied Superconductivity, 25(3), p.4200904_1 - 4200904_4, 2015/06

 Percentile:100(Engineering, Electrical & Electronic)

The Japan Atomic Energy Agency (JAEA) is responsible for procuring all amounts of Central Solenoid (CS) Conductors for ITER, including CS jacket sections. The conductor is cable-in-conduit conductor (CICC) with a central spiral. A total of 576 Nb$$_{3}$$Sn strands and 288 copper strands are cabled around the central spiral. The maximum operating current is 40 kA at magnetic field of 13 T. CS jacket section is circular in square type tube made of JK2LB, which is high manganese stainless steel with boron added. Unit length of jacket sections is 7 m and 6,300 sections will be manufactured and inspected. Outer/inner dimension and weight are 51.3/35.3 mm and around 90 kg, respectively. Eddy Current Test (ECT) and Phased Array Ultrasonic Test (PAUT) were developed for non-destructive examination. The defects on inner and outer surfaces can be detected by ECT. The defects inside jacket section can be detected by PAUT. These technology and the inspected results are reported in this paper.

JAEA Reports

Maintenance of helium refrigerator/liquefier system in ITER CS Model Coil Test Facility

Ebisawa, Noboru; Kiuchi, Shigeki*; Kikuchi, Katsumi*; Kawano, Katsumi; Isono, Takaaki

JAEA-Testing 2014-003, 37 Pages, 2015/03

JAEA-Testing-2014-003.pdf:11.7MB

Objective of the ITER CS Model Coil Test Facility is to evaluate a large scale superconducting conductor for fusion using the Central Solenoid (CS) Model Coil, which can generate a 13-T magnetic field in the inner bore with a 1.5m diameter. The facility is composed of a helium refrigerator / liquefier system, a DC power supply system, a vacuum system and a data acquisition system. This report describes that maintenance of the helium refrigerator / liquefier system since the Great East Japan Earthquake in March 2011 until the first operation after the earthquake in December 2012.

Journal Articles

Cabling technology of Nb$$_3$$Sn conductor for ITER central solenoid

Takahashi, Yoshikazu; Nabara, Yoshihiro; Ozeki, Hidemasa; Hemmi, Tsutomu; Nunoya, Yoshihiko; Isono, Takaaki; Matsui, Kunihiro; Kawano, Katsumi; Oshikiri, Masayuki; Uno, Yasuhiro; et al.

IEEE Transactions on Applied Superconductivity, 24(3), p.4802404_1 - 4802404_4, 2014/06

 Times Cited Count:16 Percentile:22.38(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency (JAEA) is procuring all amounts of Nb$$_3$$Sn conductors for Central Solenoid (CS) in the ITER project. Before start of mass-productions, the conductor should be tested to confirm superconducting performance in the SULTAN facility, Switzerland. The original design of cabling twist pitches is 45-85-145-250-450 mm, called normal twist pitch (NTP). The test results of the conductors with NTP was that current shearing temperature (Tcs) is decreasing due to electro-magnetic (EM) load cycles. On the other hand, the results of the conductors with short twist pitches (STP) of 25-45-80-150-450 mm show that the Tcs is stabilized during EM load cyclic tests. Because the conductors with STP have smaller void fraction, higher compaction ratio during cabling is required and possibility of damage on strands increases. The technology for the cables with STP was developed in Japanese cabling suppliers. The several key technologies will be described in this paper.

Journal Articles

Development of manufacturing technology of radial plate in superconducting coil for fusion reactor by diffusion bonding by Hot Isostatic Pressing (HIP)

Takano, Katsutoshi; Koizumi, Norikiyo; Masuo, Hiroshige*; Natsume, Yoshihisa*

Yosetsu Gakkai Rombunshu (Internet), 32(1), p.8 - 14, 2014/03

The authors performed trial manufacture of the RP segments using a diffusion bonding method, namely Hot Isostatic Pressing (HIP). As a result of trials, it was clarified that even when HIPping is applied, the mechanical characteristic of base metal is not deteriorated. The machining period can be reduced by half compared with the traditional manufacturing method. On the other hand, mechanical strength at 4 K is degraded due to weak bonding, that is no grain growth through joint, by HIPping. However, additional test indicates promising possibility of much better joint by higher temperature and joint surface treated HIPpings. These results justified that RP segment manufacturing is not only possible, but it is a technically valid manufacturing method that satisfies all requirements.

Journal Articles

Cable twist pitch variation in Nb$$_{3}$$Sn conductors for ITER toroidal field coils in Japan

Takahashi, Yoshikazu; Nabara, Yoshihiro; Hemmi, Tsutomu; Nunoya, Yoshihiko; Isono, Takaaki; Hamada, Kazuya; Matsui, Kunihiro; Kawano, Katsumi; Koizumi, Norikiyo; Oshikiri, Masayuki; et al.

IEEE Transactions on Applied Superconductivity, 23(3), p.4801504_1 - 4801504_4, 2013/06

 Times Cited Count:11 Percentile:35.89(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency (JAEA) is the first to start the mass production of the TF conductors in March 2010 among the 6 parties who are procuring TF conductors in the ITER project. The height and width of the TF coils are 14 m and 9 m, respectively. The conductor is cable-in-conduit conductor (CICC) with a central spiral. A circular multistage superconducting cable is inserted into a circular stainless steel jacket with a thickness of 2 mm. A total of 900 Nb$$_{3}$$Sn strands and 522 copper strands are cabled around the central spiral and the cable is inserted into a round-in-round stainless steel jacket. It was observed that the cabling pitch of the destructive sample is longer than the original pitch at cabling. The JAEA carried out the tensile tests of the cable and the measurement of the cable rotation during the insertion to investigate the cause of the elongation. The cause of elongation was clarified and the results will be described in this paper.

Journal Articles

Mass production of Nb$$_{3}$$Sn conductors for ITER toroidal field coils in Japan

Takahashi, Yoshikazu; Isono, Takaaki; Hamada, Kazuya; Nunoya, Yoshihiko; Nabara, Yoshihiro; Matsui, Kunihiro; Hemmi, Tsutomu; Kawano, Katsumi; Koizumi, Norikiyo; Oshikiri, Masayuki; et al.

IEEE Transactions on Applied Superconductivity, 22(3), p.4801904_1 - 4801904_4, 2012/06

 Times Cited Count:8 Percentile:45.33(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency is the first to start the mass production of the TF conductors in Phase IV in March 2010 among the 6 parties who are procuring TF conductors in the ITER project. The conductor is cable-in-conduit conductor with a central spiral. A total of 900 Nb$$_{3}$$Sn strands and 522 copper strands are cabled around the central spiral and then wrapped with stainless steel tape whose thickness is 0.1 mm. Approximately 60 tons of Nb$$_{3}$$Sn strands were manufactured by the two suppliers in December 2010. This amount corresponds to approximately 55% of the total contribution from Japan. Approximately 30% of the total contribution from Japan was completed as of February 2011. JAEA is manufacturing one conductor per month under a contract with two Japanese companies for strands, one company for cabling and one company for jacketing. This paper summarizes the technical developments including a high-level quality assurance. This progress is a significant step in the construction of the ITER machine.

Journal Articles

Technology development and mass production of Nb$$_{3}$$Sn conductors for ITER toroidal field coils in Japan

Takahashi, Yoshikazu; Isono, Takaaki; Hamada, Kazuya; Nunoya, Yoshihiko; Nabara, Yoshihiro; Matsui, Kunihiro; Hemmi, Tsutomu; Kawano, Katsumi; Koizumi, Norikiyo; Oshikiri, Masayuki; et al.

Nuclear Fusion, 51(11), p.113015_1 - 113015_11, 2011/11

 Times Cited Count:10 Percentile:47.37(Physics, Fluids & Plasmas)

Japan Atomic Energy Agency is procuring the Nb$$_{3}$$Sn superconductors for Toroidal Field (TF) Coils under the ITER project. Because manufacturing amount of Nb$$_{3}$$Sn strands is quite large compared with the past experience and required superconducting performance is higher than that of the model coils which have been fabricated and tested in the ITER-EDA, quality control technique is very important for the manufacture of the strands. Sophisticated control technique is also required for the jacketing, in order to fabricate the conductors with the precise outer diameter and without leakage at welding part. This paper summarizes the technical developments leading to the first successful mass production of ITER TF conductors.

Journal Articles

First qualification of ITER toroidal field coil conductor jacketing

Hamada, Kazuya; Takahashi, Yoshikazu; Isono, Takaaki; Nunoya, Yoshihiko; Matsui, Kunihiro; Kawano, Katsumi; Oshikiri, Masayuki; Tsutsumi, Fumiaki; Koizumi, Norikiyo; Nakajima, Hideo; et al.

Fusion Engineering and Design, 86(6-8), p.1506 - 1510, 2011/10

 Times Cited Count:8 Percentile:34.47(Nuclear Science & Technology)

Japan Atomic Energy Agency has a responsibility for procurement of the ITER toroidal field coil conductors as Japanese Domestic Agency (JADA) of the ITER project. The TF conductor is a circular shaped cable-in-conduit conductor, which is composed of cable and stainless steel conduit (jacket). The outer diameter and wall thickness of jacket are 43.7mm and 2mm, respectively. The cable consists of 900 Nb$$_{3}$$Sn superconducting strands and 522 Cu strands. The length of TF conductor is 780m in maximum. Preparation of conductor fabrication was completed in December 2009. And then, to demonstrate a conductor manufacturing procedure, JADA fabricated 780m-long Cu dummy conductor as a process qualification. Finally, the 780m-long Cu dummy conductor has been successfully completed, ahead of other domestic agencies that are in charge of TF conductor procurement. Since all of manufacturing processes have been qualified, JADA started to fabricate superconducting conductors for TF coils.

Journal Articles

Analytical model of the critical current of a bent Nb$$_{3}$$Sn strand

Koizumi, Norikiyo; Murakami, Haruyuki; Hemmi, Tsutomu; Nakajima, Hideo

Superconductor Science and Technology, 24(5), p.055009_1 - 055009_12, 2011/05

 Times Cited Count:18 Percentile:30.21(Physics, Applied)

Critical current performance of a large Nb$$_{3}$$Sn cable-in-conduit conductor (CICC) was degraded by periodic bending of strands due to a large transverse electromagnetic force. The degradation of each strand due to this bending should be evaluated in calculations of the critical current of a CICC, but a suitable model has not been developed yet. Therefore, the authors have developed a new analytical model which takes into account plastic deformation of copper and bronze and filament breakage. Calculated results were compared with test results for uniformly bent Nb$$_{3}$$Sn bronze-route strands. Calculated results assuming a high-transverse resistance model (HTRM) show good agreement with the test results, a finding which confirms the validity of the model. Because of a much shorter calculation time than for numerical simulation, the developed model seems much more practical for use in calculating the critical current performance of a Nb$$_{3}$$Sn CICC. In addition, simulation results show that since the neutral axis of a bent strand shifts to the compressive side due to plastic deformation of the copper and bronze, and/or filament breakage, the strand is elongated by bending. This elongation may enhance the strand's critical current performance. Moreover, calculated results indicate that dependence of the critical current on the bending strain is affected by the bending history if the strand is excessively bent, especially when filaments are broken. In a real magnet, since a strand in a CICC is normally subject to the maximum electromagnetic force prior to an evaluation of its performance at a lower electromagnetic force, the effect of over-bending should be taken into account in calculations of its critical current performance, especially when filament breakage occurs.

Journal Articles

Technology development for the manufacture of Nb$$_{3}$$Sn conductors for ITER Toroidal Field coils

Takahashi, Yoshikazu; Isono, Takaaki; Hamada, Kazuya; Nunoya, Yoshihiko; Nabara, Yoshihiro; Matsui, Kunihiro; Hemmi, Tsutomu; Kawano, Katsumi; Koizumi, Norikiyo; Oshikiri, Masayuki; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03

Japan Atomic Energy Agency is procuring the Nb$$_{3}$$Sn superconductors for Toroidal Field (TF) coils under the ITER project. Because manufacturing amount of Nb$$_{3}$$Sn strands is quite large compared with the past experience and required superconducting performance is higher than that of the model coils which have been fabricated and tested in the ITER-EDA, quality control technique is very important for the manufacture of the strands. Sophisticated control technique is also required for the jacketing, in order to fabricate the conductors with the precise outer diameter and without leakage at welding part. Cu dummy conductor with full length (760 m) has been fabricated successfully and all jacketing technology was confirmed through this fabrication. The fabrication of the Nb$$_{3}$$Sn conductor for TF coils will start in March 2010.

Journal Articles

Superconducting property and strain effect study of the Nb$$_{3}$$Sn strands developed for ITER

Nunoya, Yoshihiko; Hemmi, Tsutomu; Nabara, Yoshihiro; Matsui, Kunihiro; Isono, Takaaki; Takahashi, Yoshikazu; Koizumi, Norikiyo; Okuno, Kiyoshi

IEEE Transactions on Applied Superconductivity, 20(3), p.1443 - 1446, 2010/06

 Times Cited Count:4 Percentile:63.67(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency has developed Nb$$_{3}$$Sn strand for the ITER magnet, strand whose critical current density is about 1.4 times as large as that for ITER CS Model Coil. Magnetic field, temperature, and strain dependence on the critical current (Ic) of the strand are measured by the apparatus developed by the authors. Especially strain property is investigated in detail, and it is shown by strain tensor analysis that upper critical field dependence on strain can be naturally formulated by the high order polynomial terms of strain taking into account residual strain due to thermal contraction difference inside of the strand and applied strain externally. Correlation formula among field, temperature, strain and Ic for the strand is discussed and presented. Performance of superconducting cables composed of the developed strand is evaluated using the formula.

Journal Articles

Critical issues for the manufacture of the ITER TF coil winding pack

Koizumi, Norikiyo; Hemmi, Tsutomu; Matsui, Kunihiro; Nakajima, Hideo; Okuno, Kiyoshi; Kuno, Kazuo*; Nomoto, Kazuhiro*

Fusion Engineering and Design, 84(2-6), p.210 - 214, 2009/09

 Times Cited Count:15 Percentile:21.94(Nuclear Science & Technology)

ITER-TF coil, whose height and width are 14 m and 9 m, respectively, is scaled up by about 3 times from TF model coil (TFMC), which was developed in ITER EDA and successfully tested in EU. Although major technique of TF coil fabrication has been demonstrated in the TFMC development, new technical issues are initiated because of the scale-up. The remaining major issues are feasibilities of high accuracy automatic winding and optimization of welding of cover plates to fix the conductor in a radial plate, which is structure to enhance mechanical and electrical reliability. The authors therefore develop one of major parts of automatic winding machine, bending roller head, and successfully performed trial winding for 1/3 scale D-shaped winding. In addition, cover plate welding test was carried out using 1-m RP section and distortion of the radial plate is estimated to be in the requirement.

Journal Articles

Installation and test programme of the ITER poloidal field conductor insert (PFCI) in the ITER test facility at JAEA Naka

Nunoya, Yoshihiko; Takahashi, Yoshikazu; Hamada, Kazuya; Isono, Takaaki; Matsui, Kunihiro; Oshikiri, Masayuki; Nabara, Yoshihiro; Hemmi, Tsutomu; Nakajima, Hideo; Kawano, Katsumi; et al.

IEEE Transactions on Applied Superconductivity, 19(3), p.1492 - 1495, 2009/06

 Times Cited Count:1 Percentile:85.53(Engineering, Electrical & Electronic)

The ITER Poloidal Field Conductor Insert (PFCI) was constructed to characterize the performance of selected cable-in-conduit NbTi conductors for the ITER Poloidal Field (PF) under relevant operating conditions. The PFCI was installed and tested inside the bore of the ITER CS model coil, which provides the background magnetic field. The PFCI is a single-layer solenoid, wound from about 50 m of a full-size ITER cable-in-conduit conductor. The winding diameter and height are about 1.5 m and 1 m, respectively. The nominal design current of the conductor is 45 kA at 6 T and 5 K. The main items in the PFCI test programme are current sharing temperature (Tcs) measurements, critical current (Ic) measurements and AC loss measurement. The key technology of the installation, the test methods and procedures, and some preliminary results of the testing campaigns are described and discussed in this paper.

Journal Articles

Characterization of ITER Nb$$_{3}$$Sn strands under strain-applied conditions

Nunoya, Yoshihiko; Isono, Takaaki; Koizumi, Norikiyo; Hamada, Kazuya; Matsui, Kunihiro; Nabara, Yoshihiro; Okuno, Kiyoshi

IEEE Transactions on Applied Superconductivity, 18(2), p.1055 - 1058, 2008/06

 Times Cited Count:18 Percentile:27.72(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency has developed Nb$$_{3}$$Sn strand which can be applied to ITER TF coils. The achieved critical current density is more than 790A/mm$$^{2}$$ in bronze process strand and more than 980A/mm$$^{2}$$ in internal tin process strand under 4.2K temperature and 12T magnetic field as well as hysteresis loss of less than 770mJ/cc under $$pm$$ 3T cycle. Because these strand are utilized with an external strain by thermal contraction and hoop force of the coils, it is necessary to evaluate strain dependency of these strands to confirm the ITER conductor design. An apparatus to measure the strain dependency was newly developed. It has a horseshoe-shaped ring to produce uniform axial compressive or tensile strain along strand length and a strand is soldered on its outer surface. ITER TF coils will be operated under 0.77$$%$$ compressive strain, magnetic field of 11.3T and temperature of 5.7K as design value. Typical obtained critical current for bronze process strand is about 92A under these designed operating conditions, which is about 20$$%$$ larger than the requirement. The details of the results of strand characterization will be presented.

Journal Articles

Performance of Japanese Nb$$_{3}$$Sn conductors for ITER toroidal field coils

Takahashi, Yoshikazu; Isono, Takaaki; Koizumi, Norikiyo; Nunoya, Yoshihiko; Matsui, Kunihiro; Nabara, Yoshihiro; Hemmi, Tsutomu; Oshikiri, Masayuki; Uno, Yasuhiro*; Okuno, Kiyoshi; et al.

IEEE Transactions on Applied Superconductivity, 18(2), p.471 - 474, 2008/06

 Times Cited Count:16 Percentile:30.67(Engineering, Electrical & Electronic)

The ITER TF coil system consists of 18 D-shape coils. The operating current and the maximum field are 68 kA and 11.8 T, respectively. A Nb$$_{3}$$Sn cable-in-conduit conductor with a central channel is used, with a unit length of about 380 m. A cable consists of 900 Nb$$_{3}$$Sn strands and 522 Cu strands with a diameter of 0.82 mm. Superconducting performance of full-size conductors manufactured was measured at the operating condition of the TF coils with the maximum field. The strands made by bronze and internal-Sn methods were used for the sample conductors with a void fraction of 29% and 33%, respectively. The measured current sharing temperatures Tcs are 6.5-6.7K for the bronze route method and 5.7-5.9K for the internal-Sn method. The Tcs of the conductor with small void fraction is relatively higher with 0.1-0.2K than that with large void fraction. It is confirmed that Tcs of both strands is higher than the design value (5.7K). It is shown from the results that the strain on the conductor, estimated by the strand data, is about -0.7%. This value seems to be reasonable.

Journal Articles

Development of conduits for the ITER central solenoid conductor

Hamada, Kazuya; Nakajima, Hideo; Kawano, Katsumi; Takano, Katsutoshi; Tsutsumi, Fumiaki; Okuno, Kiyoshi; Fujitsuna, Nobuyuki*; Teshima, Osamu*

Teion Kogaku, 43(6), p.244 - 251, 2008/06

Japan Atomic Energy Agency has developed JK2LB conduit for the Nb$$_{3}$$Sn conductor of the ITER Central Solenoid. Mechanical requirements for the CS conductor conduit are 0.2% yield strength of more than 900 MPa and fracture toughness K $$_{IC}$$(J) of more than 130 MPa$$sqrt{m}$$ after a compaction and aging heat treatment (650 $$^{circ}$$C, 240 hours). In the previous work, aged JK2LB conduit has shown high strength and fracture toughness enough to satisfy the requirements. As a next step, work was performed to determine specification of the JK2LB conduit taking account of cold work including compaction and winding, and to simplify its fabrication process. To simulate the cold work effect and aging, mechanical tests were performed at 4.2 K on laboratory scale (20-30kg) ingot samples. It was found that the sum of carbon and nitrogen content should be in a range from 0.11% to 0.18% to achieve the ITER mechanical requirements. To obtain a grain size of conduit as well as that of small ingot sample, applicable solution heat treatment temperature and holding time were studied. In order to simplify the billet production process, we confirmed internal metallurgical qualities of JK2LB cast ingot. Since significant segregation was not observed, we could exclude an electroslag remelting process. Based on above achievements, full size JK2LB conduits were fabricated and satisfied the ITER mechanical requirements.

Journal Articles

How is the ITER project going on?

Takahashi, Yoshikazu

Chodendo Web 21 (Internet), (2008年1月), p.25 - 26, 2008/01

In the ITER project, it was decided in 2005 that the machine is constructed at Cadarache and the agreement of the ITER was signed by 7 parties (Japan, U.S.A., Russia, EU, Korea, China and India) in November 2006. The agreement came into effect on 24th October, 2007 and the ITER organization launched. The Japan Atomic Energy Agency was designated as the domestic agency at the same time. The status of the ITER project from 2006 to December 2007 and the Japanese contribution to the project are explained plainly.

Journal Articles

Demonstration of full scale JJ1 and 316LN fabrication for ITER TF coil structure

Hamada, Kazuya; Nakajima, Hideo; Kawano, Katsumi; Takano, Katsutoshi; Tsutsumi, Fumiaki; Okuno, Kiyoshi

Fusion Engineering and Design, 82(5-14), p.1481 - 1486, 2007/10

 Times Cited Count:8 Percentile:41.7(Nuclear Science & Technology)

no abstracts in English

56 (Records 1-20 displayed on this page)