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Summary: Less than five years after the "Atoms for Peace" speech by US President Dwight D. Eisenhower to the United Nations General Assembly in December 1953, TRIGA® (acronym for Training, Research, Isotopes, General Atomics), a new inherently safe type reactor developed for nuclear research, training and isotope production, was conceived, built and operated at the General Atomic Division in San Diego. Over the years, the TRIGA industry has soon evolved into the most widely used research reactor in the world with operating power levels up to 14 MW and designs up to 25 MW. Since 1965 the Laboratory of Applied Nuclear Energy (LENA) of the University of Pavia has been operating a TRIGA research reactor with thermal power levels of 250 kW. The installation is used to support education and training programs, neutron activation analysis activities, medical research, industrial applications, and is mainly dedicated to applied nuclear science in general. These activities will be presented together with the historical and technical aspects of the Nuclear Research Reactor., Competing Interests: The authors of this article have no conflict of interests to disclose., (Copyright© by GIMLE.)

The current Boron Neutron Capture Therapy (BNCT) experiments performed at the University of Pavia, Italy, are focusing on the in vivo irradiations of small animals (rats and mice) in order to evaluate the effectiveness of BNCT in the treatment of diffused lung tumors. After the irradiation, the animals are manipulated, which requires an evaluation of the residual radioactivity induced by neutron activation and the relative radiological risk assessment to guarantee the radiation protection of the workers. The induced activity in the irradiated animals was measured by high-resolution open geometry gamma spectroscopy and compared with values obtained by Monte Carlo simulation. After an irradiation time of 15 min in a position where the in-air thermal flux is about 1.2 × 10(10) cm(-2) s(-1), the specific activity induced in the body of the animal is mainly due to 24Na, 38Cl, 42K, 56Mn, 27Mg and 49Ca; it is approximately 540 Bq g(-1) in the rat and around 2,050 Bq g(-1) in the mouse. During the irradiation, the animal body (except the lung region) is housed in a 95% enriched 6Li shield; the primary radioisotopes produced inside the shield by the neutron irradiation are 3H by the 6Li capture reaction and 18F by the reaction sequence 6Li(n,α)3H → 16O(t,n)18F. The specific activities of these products are 3.3 kBq g(-1) and 880 Bq g(-1), respectively.

Compared to conventional engineering systems, nuclear reactors present unique physical and safety features that make their high-fidelity modelling both necessary and complex. Indeed, modelling nuclear reactors is intrinsically a multi-scale and multi-physics task. Up to recent years, the modelling approach for nuclear reactors saw the use of highly performing computer codes and hardware to retrieve a model as close as possible to reality; whereas this remains true, especially from the point of view of regulators and safety assessments, an alternative modelling approach has appeared in the nuclear reactor world. Indeed, whereas high-fidelity models are invaluable for providing in-depth insights into the system, especially when experimental data are not available, their computational cost is such that they are not suited for all applications that involve multiple simulations over a parametric domain (such as in the design and optimisation phase, known as multi-query scenarios). Thus, this novel approach aims at reducing the computational complexity of high-fidelity models whilst preserving high enough accuracy to satisfy the regulatory requirements of the nuclear world, and modelling techniques with this scope fall into the broad category of Model Order Reduction (MOR) techniques. MOR methodologies offer a trade-off between computational cost and solution accuracy; they can also jointly work with Data Assimilation techniques, which deal with the dynamic integration of experimental data and numerical estimates, thus surpassing the logic of using experimental data only as a posteriori validation tool. As the use of MOR and Data Assimilation (DA) for nuclear reactor analysis and for the development of integrated tools and digital twins for the system is still in the first stages, this work overviews some MOR and DA methodologies developed by the authors applied to an existing nuclear system, the TRIGA Mark II research reactor at the University of Pavia, which represents a benchmark test case of a complete nuclear reactor with experimental data available. The choice of using different MOR techniques to tackle various problems follows the logic of developing specific algorithms for specific issues and then merging them into a single MOR and DA-based digital twin, thus reducing the complexity and the cost of the single algorithm modules compared to a single general one: the three MOR techniques considered in this work (Dynamic Mode Decomposition, Proper Orthogonal Decomposition with Kalman Filtering and Generalised Empirical Interpolation Method) follows this logic. Indeed, the results show the potentiality of this approach for complex engineering problems, showing how these techniques can offer significant insights into the system without the computational cost associated with high-fidelity models. • Definition of the MOR and DA frameworks for reactor dynamics analysis. • Three different MOR techniques have been considered: DMD, POD with KF nd GEIM. • Application to the TRIGA Mark II reactor of the University of Pavia as a test case. • Good results in state prediction and model improvement. [ABSTRACT FROM AUTHOR]

A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyze neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate low power experimental reactors from those used for power production, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the core and to obtain a substantial increase in the Core Excess reactivity value. The evaluation of fuel burnup and the reconfiguration results are presented in this paper. [ABSTRACT FROM AUTHOR]

In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics. [ABSTRACT FROM AUTHOR]