SUMRAY: R and Python codes for calculating cancer risk due to radiation exposure of a population

Sasaki, Michiya*; Furukawa, Kyoji*; Satoh, Daiki; Shimada, Kazumasa; Kudo, Shinichi*; Takagi, Shunji*; Takahara, Shogo; Kai, Michiaki*

Journal of Radiation Protection and Research, 48(2), p.90 - 99, 2023/06

https://doi.org/10.14407/jrpr.2022.00213

This paper reports on the calculation code that is the result of the activities of the "Task Group for Development of Cancer Risk Estimation Codes Associated with Radiation Exposure (FY2020-2021)" established by the Japan Health Physics Society. In order to promote research on the estimation of cancer risk associated with radiation exposure, the Task Group decided to disclose the source code, including the algorithm and parameters used in the calculations, and to release the code under a license that permits modification and redistribution of the code. The computational code was named SUMRAY and coded in two computer languages, that is R and Python. The code is capable of calculating the accumulated excess risk using Monte Carlo methods with a 95% confidence interval. The results of SUMRAY were compared with the results of the existing codes whose source code is not publicly available, under the same calculation conditions. From the results, it was found that they were in reasonable agreement within the confidence interval. It is expected that SUMRAY, an open-source software, will be used as a common basis for cancer risk estimation studies associated with radiation exposure.

Liquid divertor

Shimada, Michiya; Miyazawa, Junichi*

Purazuma, Kaku Yugo Gakkai-Shi, 92(2), p.119 - 124, 2016/02

AA2015-0751.pdf:0.61MB

http://www.jspf.or.jp/Journal/PDF_JSPF/jspf2016_02/jspf2016_02-119.pdf

Actively convected liquid metal divertor is promising for providing a solution for issues of DEMO reactors including heat removal and disruptions. This chapter gives an overview of the motivation, research history, recent development, future perspective and issues to be resolved.

Actively convected liquid metal divertor

Shimada, Michiya; Hirooka, Yoshihiko*

Nuclear Fusion, 54(12), p.122002_1 - 122002_7, 2014/12

https://doi.org/10.1088/0029-5515/54/12/122002

Tungsten is considered to be the most promising material for divertor in a fusion reactor. Tungsten divertor can withstand the heat loads of ITER, but the heat loads of DEMO divertor is a challenge. Pulsive heat loads as those associated with disruption could melt tungsten targets. The surface would not be flat after subsequent resolidification, which would significantly deteriorate its heat handling capability. Furthermore, DBTT of tungsten is rather high: 400C, which would become even higher after neutron irradiation, possibly resulting in cracks in tungsten. Our proposal is to use liquid metal for the divertor target material and actively circulate it with force. A simplified analysis of mhd equation in a cylindrical geometry suggests that the engineering requirement is modest. This analysis suggests that this new divertor concept merits further investigation.

ITER vacuum vessel, in-vessel components and plasma facing materials

Ioki, Kimihiro*; Barabash, V.*; Cordier, J.*; Enoeda, Mikio; Federici, G.*; Kim, B. C.*; Mazul, I.*; Merola, M.*; Morimoto, Masaaki*; Nakahira, Masataka*; et al.

Fusion Engineering and Design, 83(7-9), p.787 - 794, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.05.027

This paper presents recent results of ITER activities on Vacuum Vessel (VV), blanket, limiter, and divertor. Major results can be summarized as follows. (1) The VV design is being developed in more details considering manufacturing and assembly methods, and cost. Incorporating manufacturing studies being performed in cooperation with parties, the regular VV sector design has been nearly finalized. (2) The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. (3) The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule.

Progress in the ITER physics basis, 1; Overview and summary

Shimada, Michiya; Campbell, D. J.*; Mukhovatov, V.*; Fujiwara, Masami*; Kirneva, N.*; Lackner, K.*; Nagami, Masayuki; Pustovitov, V. D.*; Uckan, N.*; Wesley, J.*; et al.

Nuclear Fusion, 47(6), p.S1 - S17, 2007/06

https://doi.org/10.1088/0029-5515/47/6/S01

The Progress in the ITER Physics Basis document is an update of the ITER Physics Basis (IPB), which was published in 1999. The IPB provided methodologies for projecting the performance of burning plasmas, developed largely through coordinated experimental, modeling and theoretical activities carried out on today's tokamaks (ITER Physics R&D). In the IPB, projections for ITER (1998 Design) were also presented. The IPB also pointed out some outstanding issues. These issues have been addressed by the International Tokamak Physics Activities (ITPA), which were initiated by the European Union, Japan, Russia and the U.S.A.. The new methodologies of projection and control developed through the ITPA are applied to ITER, which was redesigned under revised technical objectives, but will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy.

Progress in the ITER physics basis, 8; Plasma operation and control

Gribov, Y.*; Humphreys, D. A.*; Kajiwara, Ken*; Lazarus, E. A.*; Lister, J. B.*; Ozeki, Takahisa; Portone, A.*; Shimada, Michiya*; Sips, A. C. C.*; Wesley, J. C.*

Nuclear Fusion, 47(6), p.S385 - S403, 2007/06

https://doi.org/10.1088/0029-5515/47/6/S08

This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This chapter considers only plasma initiation and plasma basic control. The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, ITER can produce plasma initiation in a low toroidal electric field of 0.3V/m, if it is assisted by about 2MW of ECRF heating. The plasma basic control is described, which includes control of the plasma current, position and shape - the plasma magnetic control, as well as control of other plasma global parameters or their profiles - the plasma performance control.

Study on current drive capability of lower hybrid waves and neutral beam in an ITER steady state scenario

Oikawa, Toshihiro; Shimada, Michiya; Polevoi, A. R.*; Naito, Osamu; Bonoli, P. T.*; Hayashi, Nobuhiko; Kessel, C. E.*; Ozeki, Takahisa

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 7 Pages, 2007/03

http://www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/node75.htm#19564

The current drive capability of lower hybrid waves is assessed for an ITER steady state scenario using a relativistic, one-dimensional Fokker-Planck code and a toroidal ray tracing code. The present LH launcher design provides a current drive efficiency of 1.810AWm and an off-axis profile of the driven current that is fovarable for a reversed magnetic shear configuration. Possible optimizations in the LH power spectrum are investigated. Neutral beam current drive (NBCD) is investigated with theoretical codes employing different approaches. An NBCD code employing the bounce-averaged Fokker-Planck equation to include orbit effects in a toroidal system shows a good agreement with an orbit following Monte-Carlo code.

Disruption scenarios, their mitigation and operation window in ITER

Shimada, Michiya; Sugihara, Masayoshi; Fujieda, Hirobumi*; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03

http://www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/node89.htm

Several representative disruption scenarios are specified and disruption simulations are performed with the DINA code and EM load analyses with the 3D FEM code for these scenarios based on newly derived physics guidelines. Although some margin is confirmed in the EM loads due to induced eddy and halo currents on the in-vessel components for all of the representative scenarios, but the margin is not large. The heat load on various parts of the first wall due to vertical movements and thermal quenches is calculated. The beryllium wall will not melt during vertical movement. Melting is anticipated at the thermal quench during a VDE, though its impact could be reduced substantially by implementing a reliable detection and mitigation system, e.g., massive gas injection. With unmitigated disruptions, the loss of beryllium layer is expected to be within 30 m/event out of 10 mm thick beryllium first wall.

Fusion plasma research toward fusion power plants

Kamada, Yutaka; Shimada, Michiya; Miura, Yukitoshi; Ogawa, Yuichi*

Purazuma, Kaku Yugo Gakkai-Shi, 81(11), p.849 - 862, 2005/11

https://doi.org/10.1585/jspf.81.849

This section discusses fusion plasma research that needs to be carried out to develop fusion power plants. Burning plasma, in which self-heating by energetic alpha particles plays an essential role, should be recognized as autonomous system. This is quite different from present plasma experiments, suggesting a possibility to yield some qualitative changes in fusion plasma research. Research with ITER is strongly expected to contribute to this burning plasma physics. In addition, plasma performance in steady-state and at high beta is very important in fusion power plants from the engineering and economical viewpoints. Plasma parameters expected for fusion power plants are discussed, and present status of experimental research is reviewed. Research in devices other than ITER with unique features would be instrumental for exploring high performance plasmas. A necessity of research complementary to ITER plasma is discussed.

Progress in physics basis and its impact on ITER

Shimada, Michiya; Campbell, D.*; Stambaugh, R.*; Polevoi, A. R.*; Mukhovatov, V.*; Asakura, Nobuyuki; Costley, A. E.*; Donn, A. J. H.*; Doyle, E. J.*; Federici, G.*; et al.

Proceedings of 20th IAEA Fusion Energy Conference (FEC 2004) (CD-ROM), 8 Pages, 2004/11

This paper summarises recent progress in the physics basis and its impact on the expected performance of ITER. Significant progress has been made in many outstanding issues and in the development of hybrid and steady state operation scenarios, leading to increased confidence of achieving ITER's goals. Experiments show that tailoring the current profile can improve confinement over the standard H-mode and allow an increase in beta up to the no-wall limit at safety factors 4. Extrapolation to ITER suggests that at the reduced plasma current of 12MA, high Q 10 and long pulse (1000 s) operation is possible with benign ELMs. Analysis of disruption scenarios has been performed based on guidelines on current quench rates and halo currents, derived from the experimental database. With conservative assumptions, estimated electromagnetic forces on the in-vessel components are below the design target values, confirming the robustness of the ITER design against disruption forces.