Reactor Pulse Operation for Nuclear Instrumentation Detector Testing – Preparation of a Dedicated Experimental Campaign at the JSI TRIGA Reactor

The availability of neutron fields with a high neutron flux, suitable for irradiation testing of nuclear instrumentation detectors relevant for applications in nuclear facilities such as material testing reactors (MTRs), nuclear power reactors and future fusion reactors is becoming increasingly limited. Over the last several years there has been increased interest in the experimental capabilities of the 250 kW Jožef Stefan Institute (JSI) TRIGA research reactor for such applications, however, the maximal achievable neutron flux in steady-state operation mode falls short of MTR-relevant conditions. The JSI TRIGA reactor can also operate in pulse mode, with a maximal achievable peak power of approximately 1 GW, for a duration of a few ms. A collaboration project between the JSI and the French Atomic and Alternative Energy Commission (CEA) was initiated to investigate absolute neutron flux measurements at very high neutron flux levels in reactor pulse operation. Such measurements will be made possible by special CEA-developed miniature fission chambers and modern data acquisition systems, supported by the JSI TRIGA instrumentation and activation dosimetry. Additionally, measurements of the intensity of Cherenkov light are proposed and being investigated as an alternative experimental method. This paper presents the preparatory activities for an exhaustive experimental campaign, which were carried out in 2019-2020, consisting of test measurements with not fully appropriate fission chambers, activation dosimetry and silicon photomultipliers (SiPMs) The presented results provide useful and promising experimental indications relevant for the design of the experimental campaign.

Reactor with metallic fuel and lead-208 coolant

The paper considers the concept of a fast lead cooled 25MW reactor for a variety of applications, including incineration of minor actinides, production of medical radioisotopes, testing of radiation-damaged nuclear technology materials, etc. A specific feature of the proposed reactor is rather a high neutron flux of 2.6·1015 n/(cm2·s) at the core center, high average neutron energy of 0.95 MeV at the core center, and a large fraction (40%) of hard neutrons (En > 0.8 MeV). The extremely high estimated reactor parameters are achieved thanks to the small core dimensions (DxH ≈ 0.50×0.42 m2), innovative metallic fuel of the Pu-Am-Np-Zr alloy, and the 208Pb enriched lead coolant. A relatively high probability of 241Am fission (about 50%) is achieved in the reactor core’s hard spectrum, this making it possible to incinerate up to 4 kg of 241Am during one reactor campaign of 1000 effective days.

The DN-6 Neutron Diffractometer for High-Pressure Research at Half a Megabar Scale

A neutron diffractometer DN-6 at the IBR-2 high-flux reactor is used for the studies of crystal and magnetic structure of powder materials under high pressure in a wide temperature range. The high neutron flux on the sample due to a parabolic focusing section of a neutron guide and wide solid angle of the detector system enables neutron diffraction experiments with extraordinarily small volumes (about 0.01 mm3) of studied samples. The diffractometer is equipped with high-pressure cells with sapphire and diamond anvils, which allow pressures of up to 50 GPa to be reached. The technical design, main parameters and current capabilities of the diffractometer are described. A brief overview of recently obtained results is given.

Stress Evaluation Method of Baffle Former Bolt and its Maintenance Program

Baffle Former Bolt (BFB) is a fastening part of Reactor Vessel Internals (RVI) of PWR. BFB is made of type 347 or 316CW (cold work) stainless steel and it is known to have the risk of cracking caused by Irradiation Assisted Stress Corrosion Cracking (IASCC) under high neutron flux and tensile stress. To evaluate the time to crack of BFB, BFB’s time-dependent stress change caused by irradiation creep (relaxation) and by the swelling deformation of a baffle structure should be obtained. The authors have developed the finite element (FE) analysis method to calculate time-dependent stress of BFB considering the irradiation effects. The method combines two kinds of models; “global model” to calculate the deformation of whole baffle structure and “local model” to calculate the peak stress at the stress concentrated area under the bolt head. Incorporating the above calculation method, a new BFB inspection and evaluation guideline has been established in Japan. The concept of the guideline is also outlined in the paper.