Development of in-vessel neutron flux monitor equipped with microfission chambers to withstand the extreme ITER environment

Ishikawa, Masao; Takeda, Keigo; Itami, Kiyoshi

Fusion Engineering and Design, 109-111(Part B), p.1399 - 1403, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.12.012

The in-vessel neutron flux monitor equipped with Microfission Chambers (MFCs) has been developed. Since in-vessel components of the MFC are exposed to the extreme ITER environment, such as high radiation, high temperature and high electromagnetic (EM) forces, those components need to withstand such ITER environment. In this study, the in-vessel components of the MFC have been developed in order to apply ITER conditions. For these components, several cases of analyses (neutronic, thermal and EM analysis) and soundness verification tests (high-temperature, vibration and noise immunity test) have been conducted. As a result, it is demonstrated that the in-vessel components of the MFC can be used in the extreme ITER environments for 20 years without any replacements.

Mechanical properties of F82H plates with different thicknesses

Sakasegawa, Hideo; Tanigawa, Hiroyasu

Fusion Engineering and Design, 109-111(Part B), p.1724 - 1727, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.10.017

Fusion DEMO reactor requires over 11,000 tons of reduced activation ferritic/martensitic steel and it is important to develop the manufacturing technology for producing large-scale components of DEMO blanket with appropriate mechanical properties. In this work, we studied mechanical properties of ferritic/martensitic steel F82H plates with different thicknesses. This is because mechanical properties are generally degraded with increasing production volume and size. As the result, their homogeneity and anisotropy were not significant. However, mass effect was found in their Charpy impact property with increasing plate thickness, i.e. the ductile brittle transition temperature (DBTT) of a 100 mm thick plate was higher than those of the other plates, but its DBTT was still lower than 0C and comparable to the former heats.

New remarks on KERMA factors and DPA cross section data in ACE files

Konno, Chikara; Sato, Satoshi; Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro

Fusion Engineering and Design, 109-111(Part B), p.1649 - 1652, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.10.038

Recently we have examined KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 in more detail and we found out the following new problems on the KERMA factors and DPA cross section data. (1) NJOY bugs and incorrect nuclear data generated KERMA factors and DPA cross section data of no increase with decreasing neutron energy in low neutron energy. (2) Huge helium production data caused drastically large KERMA factors and DPA cross section data in low neutron energy. (3) It seemed that NJOY could not adequately process capture cross section data in File 6, not File 12-15. (4) KERMA factors with the kinematics method are not correct for nuclear data libraries without detailed secondary particle data (energy-angular distribution data). These problems should be resolved based on our study.

Development of a dummy load and waveguide components for 1 MW CW gyrotron

Ioki, Kimihiro*; Hiranai, Shinichi; Moriyama, Shinichi; Tanaka, Suguru*

Fusion Engineering and Design, 109-111(Part A), p.951 - 955, 2016/11

https://doi.org/10.1016/j.fusengdes.2016.01.046

A dummy load dissipates the RF power and is required to test and adjust a gyrotron or a transmission line. The most critical issue is long-term reliability in the vacuum and coolant boundary of the rotation mechanism for current large-scale dummy loads. A new design has been developed to use linear movement for the reflector assembly to mitigate the heat deposition concentration. The thickness distribution of the ceramic layer is carefully analyzed and optimized. A prototypical dummy load will be manufactured as the next step.

Evaluation of impacts of stress triaxiality on plastic deformability of RAFM steel using various types of tensile specimen

Kato, Taichiro; Ohata, Mitsuru*; Nogami, Shuhei*; Tanigawa, Hiroyasu

Fusion Engineering and Design, 109-111(Part B), p.1631 - 1636, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.11.003

Plastic deformability of material shows general tendency to decrease due to become hard and brittle. Also, the plastic deformability tends to decrease as the stress triaxiality of a parameter to evaluate the magnitude of plastic constraint increases. Therefore, it is necessary to accurately understand the ductility loss limit of RAFM in order to conduct the structural design assessment of a fusion reactor demo blanket. In this study, plastic deformability of RAFM was evaluated the impacts of stress triaxiality on variation of tensile specimen shape and testing conditions. In the results, the fracture was defined as not the point of macro-crack but that of micro-crack. It was confirmed that the true strain rate significantly increases in the vicinity of the point of micro-crack. The relationships between the fracture ductile and stress triaxiality of the full size tensile specimen and the miniature size one were shown on the single curve regardless of the specimen size.

A New blanket tritium recovery experiment with intense DT neutron source at JAEA/FNS

Ochiai, Kentaro; Edao, Yuki; Hoshino, Tsuyoshi; Kawamura, Yoshinori; Ota, Masayuki; Kwon, Saerom; Konno, Chikara

Fusion Engineering and Design, 109-111(Part B), p.1143 - 1147, 2016/11

https://doi.org/10.1016/j.fusengdes.2016.01.002

We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the more detail tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that the further improvement on the quantitative tritium measurement by IC method was needed.

Development of manufacturing technology for ITER TF coil structure

Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Inagaki, Takashi; Matsui, Kunihiro; Koizumi, Norikiyo

Fusion Engineering and Design, 109-111(Part B), p.1592 - 1597, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.11.013

Japan Atomic Energy Agency (JAEA) has responsibility to procure 9 Toroidal Field (TF) coils and 19 TF coil structures for ITER. A TF coil structure consists of the main body structure having a D-shape with 16.5 m in height and 9m in width in which superconducting winding is stored and the components to connect adjacent TF coil or other ITER devices. TF coil structures are required the very tight tolerance which is less than 2 mm for the final dimension, which is quite challenging considering large size of TF coil structure. To achieve this tolerance, extra material will be put on the each material, and machining must be performed after welding. It is important to figure out detail welding deformation and reducing the machining process to optimize manufacturing. JAEA performed an additional manufacturing trial of A1 segment which is part of TF coil structure. JAEA adopted balance welding instead of using strong restriction jig welding in additional trial. The angular distortion of previous result was +6.5/+8.9mm, however angular distortions of latest trial were -3.0/+1.6mm (right side) and 0.0/+2.4mm (left side). This progress shows that welding deformation could be controlled closer in the target value (0.0 mm) than previous method applied. Based on latest knowledge, JAEA started actual TF coil structure manufacturing from April 2014. Actual manufacturing is steadily progressing with development process improvement by learning effect and improvement of manufacturing sequence.

A New integral experiment on copper with DT neutron source at JAEA/FNS

Kwon, Saerom; Sato, Satoshi; Ota, Masayuki; Ochiai, Kentaro; Konno, Chikara

Fusion Engineering and Design, 109-111(Part B), p.1658 - 1662, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.10.036

A benchmark experiment on copper with the DT neutron source at JAEA/FNS was performed over 20 years ago. The experiment showed that ratios of the calculated values to the experimental ones (C/Es) related to lower energy neutrons were drastically smaller than unity. In order to reveal reasons of the small C/Es, we performed an integral experiment on copper with the DT neutron source at JAEA/FNS. A quasi-cylindrical copper assembly of 315 mm in radius and 608 mm in depth was covered with LiO blocks of 51 mm in thickness for the front and side parts and 153 mm in thickness for the rear part to exclude background neutrons. We measured the reaction rates of 5 reactions and the fission rates of U and U. The experiment was analyzed by using MCNP5-1.40 with the recent nuclear data libraries, ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0. As a result, the C/E of the reaction rate of the Au(n,)Au reaction was improved by 10% from the previous result and the combination of the Cu data in JEFF-3.2 and Cu data in JENDL-4.0 increased the C/E by more 10%. Moreover, the calculated result with modified elastic scattering and capture cross section data by 10% resolved the underestimation. The copper data should be reevaluated.

Synthesis and characteristics of ternary Be-Ti-V beryllide pebbles as advanced neutron multipliers

Kim, Jae-Hwan; Nakamichi, Masaru

Fusion Engineering and Design, 109-111(Part B), p.1764 - 1768, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.09.015

Safety studies of plasma-wall events with AINA code for Japanese DEMO

Rivas, J. C.*; Nakamura, Makoto; Someya, Yoji; Takase, Haruhiko; Tobita, Kenji; Dies, J.*; Blas, A. de*; Fabbri, M.*; Riego, A.*

no journal, , 

In the frame of JAPAN-EU collaborative work for development of AINA code in 2014-2016, a version of AINA code has been developed for the Japanese DEMO WCPB design. During 2014, the AINA code was adapted from ITER to this new mission. A breeding blanket model was implemented in code. The configuration was changed to implement the design parameters of DEMO reactor. Finally, safety studies of plasma-wall transients affecting blanket region were performed. During 2015, plasma models were improved both for plasma core and for divertor (improved SOL model). Safety analyses affecting divertor were performed, considering thermohydraulic accidents and plasma transients where loss of control function was assumed. First analyses performed for the Japanese DEMO design show the behavior of the reactor during Ex-Vessel LOCA and during overpower events. The preliminary conclusions point to the possibility of considering the plasma control system as a safety important component.