The work presented in this thesis deals with the characterization of gamma field inside a nuclear reactor by experiments and computational modelling. In the first part of the thesis an outline of the nuclear with neutrons and neutron transport. A description of high energy photon and electron reactions and importance of their coupling for accurate calculations of energy deposition. Particle transport equations are presented with emphasis on deriving adjoint operators used for variance reduction of Monte Carlo particle transport codes. Characterization of gamma radiation field using Monte Carlo transport codes only takes into account prompt gamma generation from fission, inelastic scattering and prompt (n,gamma) reactions. Previous evaluations suggest a roughly 30 % underestimation compared to measurements. A JSIR2S code package for delayed radiation field calculations has been developed and validated by numerous experiments. Characterization of neutron and prompt gamma radiation field inside the JSI TRIGA reactor core irradiation facilities was performed using the kerma approximation. The computational model was later expanded and the criticality source term translated to a fixed source for calculations of variance reduction parameters. The methodology has been validated by experiments, showing good agreement for neutrons, while underestimating the gamma field due to neglecting delayed radiation field. Several experimental campaigns were performed at JSI TRIGA reactor using fission and ionization chamber and Thermoluminescent dosimeters. An experimental procedure for estimation of the delayed gamma fraction was developed. Validation of the JSIR2S was performed on the above mentioned measurements, showing agreement within the uncertainty. use case on using the JSIR2S for calibration of semiconductor detectors in the JSI TRIGA reactor after reactor shut-down is described. The JSIR2S code package is also applied to shut-down dose rate calculations in fusion problems showing good agreement with experiments and similar two-step and single-step methodology codes for delayed radiation field characterization.
D.09 Tutoring for postgraduate students
COBISS.SI-ID: 16825859This thesis describes the development of a stochastic method to calculate gamma and neutron dose rates for the JSI TRIGA reactor, which can be applied to normal and emergency operations. Knowing the dose rates during normal operation is essential to keep workers safe and in the design of appropriate shielding for new experiments while knowing the dose rates during a postulated accident scenario is necessary for developing an emergency response plan. A completely new MCNP model was designed containing the reactor core and the surrounding components within the concrete shield. Furthermore, the reactor platform, reactor hall, reactor basement and the control room were included in order to calculate the gamma and neutron dose fields within the radiation controlled area. Neutrons and prompt gamma rays were validated in the case where a beam tube was left open, and the reactor was at low power. In this way, neutron and gamma dose rate measurements could be taken around the beam port. The delayed gamma source was validated in the case where one of the irradiated fuel elements was placed in a transport cask, and the surrounding dose rates measured. The good agreement between the calculated and measured results meant the model could be used to predict dose rates during normal operation with newly designed shielding for the beam tube no. 5. Before the shield was constructed, its performance was evaluated by the methodology developed in this thesis. Furthermore, an accident scenario involving the loss of water (LOWE) in the reactor pool and the spent fuel pool were analysed using the same methodology. The LOWE scenario is one of several design-based accidents scenarios to be considered when operating the JSI TRIGA reactor, that was analysed for the first time by the same method. The LOWE for the reactor pool was previously analysed using deterministic methods. Provided results will be used for the next revision of the Safety Analysis Report of the JSI TRIGA Mark II research reactor.
D.09 Tutoring for postgraduate students
COBISS.SI-ID: 49249795NENE 2020 [Elektronski vir] : 29th International Conference Nuclear Energy for New Europe : September 7-10, Portorož, Slovenia : NENE 2020 conference proceedings International Conference Nuclear Energy for New Europe (29 ; 2020 ; Portorož)
B.02 Presiding over the programming board of a conference
COBISS.SI-ID: 37460227The JSI TRIGA reactor features several in-core and ex-core irradiation facilities, each having different properties, such as neutron/gamma flux intensity, spectra and irradiation volume. A series of experiments and calculations was performed in order to characterise radiation fields in irradiation channel thus allowing users to perform irradiations in a well characterised environment. Since 2001 the reactor has been heavily used for radiation hardness studies for components used at accelerators such as the Large Hadron Collider (LHC) at CERN. Since 2010 it has been extensively used for testing of new detectors and innovative data acquisition systems and methods developed and used by the CEA. Recently, several campaigns were initiated to characterise the gamma field in the reactor and use the experimental data for improvement of the treatment of delayed gammas in Monte Carlo particle transport codes. In the future it is planned to extend the testing options by employing pulse mode operation, installation of a high energy gamma ray irradiation facility and allow irradiation of larger samples at elevated temperature.
B.04 Guest lecture
COBISS.SI-ID: 13466115