JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described.

  The Particle and Heavy Ion Transport code System, PHITS, has been developed under the collaboration of several research institutes in Japan and Europe. This system can simulate the transport of most particles with energy levels up to 1 TeV (per nucleon for ion) using different nuclear reaction models and data libraries. More than 2,500 registered researchers and technicians have used this system for various applications such as accelerator design, radiation shielding and protection, medical physics, and space- and geo-sciences. This paper summarizes the physics models and functions recently implemented in PHITS, between versions 2.52 and 2.88.

  Geant4 is a software toolkit for simulating interactions between particles and matter. The Geant4 collaboration develops and maintains it. The Japanese group played a very important role even before the beginning of the collaboration. The history, unwritten background information and the recent status of Geant4 are described in the Japanese developers’ point of view.

  Recently, the development of the Monte Carlo calculation technology in the field of radiation medicine is remarkable. The Monte Carlo method is used for the dose calculation of treatment planning in the field of the radiation therapy. And some dose evaluation tools using Monte Carlo method have been already developed in the field of the radiation diagnostics. Here we report about clinical situation and research situation on the use of Monte Carlo calculation in the recent radiation medicine field.

  In the shielding calculation of the nuclear fusion reactor, neutron and photon transport calculations have been performed by using three dimensional Monte Carlo calculation code MCNP. Recent applications with MCNP are reported in the shielding calculations of the nuclear fusion reactor. CAD/MCNP conversion codes have been developed, and MCNP calculation geometry data are automatically created by CAD/MCNP conversion codes. Details and application results are reported on the CAD/MCNP conversion code. By modifying the MCNP code and nuclear data library, the effective dose rates after reactor shutdown have been calculated with high accuracy. The calculation method and the application results are reported on the effective dose rates after shutdown.