OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths.

Occupational radiation exposure during removal of radioactive reactor components from GRR-1 pool.

The aim of the study was to control occupational exposure during the removal of radioactive reactor components from a Greek research reactor pool. The method comprised the prediction of the radiation levels, the design of special shielding structures and the occupational dose assessment. Activation calculations were performed using the FISPACT code to predict the source term.

Application of the Monte Carlo method for the calibration of an in situ gamma spectrometer.

A MCNP model was developed for the efficiency calibration of an in situ gamma ray spectrometry system based on a high purity germanium (HPGe) detector. The detector active crystal volume was adjusted semi-empirically against experimental measurements. Calculated full energy peak efficiency curves, over the photon energy range between 50 keV and 5 MeV, are presented for surface and slab source configurations.

An MCNP-based model for the evaluation of the photoneutron dose in high energy medical electron accelerators.

The development of a computational model for the treatment head of a medical electron accelerator (Elekta/Philips SL-18) by the Monte Carlo code mcnp-4C2 is discussed. The model includes the major components of the accelerator head and a pmma phantom representing the patient body. Calculations were performed for a 14 MeV electron beam impinging on the accelerator target and a 10 cmx10 cm beam area at the isocentre.

Nondestructive characterization of radioactive waste drums by gamma spectrometry: a Monte Carlo technique for efficiency calibration.

A semi-empirical non-destructive technique to assay radioactive waste drums is presented. The technique is based on gamma spectrometry performed using a portable NaI detector and Monte Carlo simulations using the MCNP code in order to derive the gamma ray detector efficiency for the volume source. The derivation of detector efficiency was performed assuming homogeneous distribution of the source activity within the matrix material.