The 3 MW TRIGA (Training, Research, Isotopes, General Atomics) Philippine Research Reactor-1 (PRR-1) at the DOST-PNRI achieved its first criticality on 08 Mar 1988 after its successful upgrade from a plate-type reactor. However, due to unresolved technical problems discovered weeks after the upgrade, the PRR-1 was considered inoperable and has been in shutdown status since then. The slightly irradiated TRIGA fuel rods of the PRR-1 are currently in an interim storage tank and are planned to be utilized in a subcritical reactor assembly. As part of the project to reuse the fuels, simulation models for both present and proposed configurations are important. In this work, we present the complete model of the former configuration of PRR-1 with 115 TRIGA fuel rods developed with the Serpent Monte Carlo code version 2 for simulation of criticality and neutronic analysis. The model of the TRIGA fuel rods was validated in the fresh fuel configuration through the benchmark analysis described in the 1988 reactor criticality report. The effective multiplication factors from the Serpent-2 simulation (𝒌𝒆𝒇𝒇 = 1.0690 ± 0.0012) and measured value of 1.0661 have been found to agree with a deviation of 259 pcm. Neutron flux and fission power distribution simulations using the same reactor configuration were also presented to serve as reference for future burn up calculations and fuel characterization.

The Philippine Research Reactor-1 (PRR-1) will be revived as a subcritical assembly for training, education, and research (SATER) in the field of nuclear science and technology. SATER will utilize the existing slightly irradiated nuclear fuel rods that have been maintained in the PRR-1 facility for more than 30 years. A subcritical arrangement for the fuel rods was chosen considering the inherent safety of this configuration. In this paper, we calculate relevant reactor parameters that characterize different annular and hexagonal subcritical configurations of 44 PRR-1 fuel rods. These parameters include the neutron multiplication factor ( ), the effective delayed neutron fraction ( ), and the mean neutron generation time ( ), which are essential quantities to describe reactor behavior. Calculations were performed using the well-validated Monte Carlo radiation transport code MCNP5 v.1.6 together with the ENDF/B- VII.1 evaluated nuclear data library. The maximum value is at 4.0 cm pitch for the chosen annular arrangement, while the maximum value is at 4.3 cm pitch for the chosen hexagonal arrangement. For these configurations, the reactor kinetic parameters were and for the annular arrangement, while and for the hexagonal arrangement. Results demonstrate that with 44 fuel rods, different fuel arrangements remain subcritical with a subcriticality margin that is at least or maximum of 0.97. The key reactor performance characteristics determined in this study can aid in the analysis of transient behavior and safety assessment of subcritical core configurations with TRIGA fuel rods. Our results provide support in expanding the utilization options for irradiated TRIGA fuel rods, even for other TRIGA facilities

The Philippine Research Reactor-1 (PRR-1) fuel storage facility is a wet storage for irradiated nuclear fuel. The behavior of the facility in case of accidental loss of water is currently unknown. Safety analysis of a repository of fissile material involves criticality safety assessment and radiation dose estimation for normal and accident scenarios. To determine the radiological consequences of loss of water inventory in the fuel storage facility, we calculated the effective multiplication factor (keff) and gamma dose rate distribution in the system using Monte Carlo N-particle (MCNP) radiation transport code. Results show that keff will decrease from a maximum of 0.61424 ± 0.00013 when the water level is decreased from its fully moderated condition. In contrast, keff will increase to a maximum of 0.7468 ± 0.0002 when voids are introduced in the water. These results indicate that the system will remain subcritical (keff<1) and a runaway supercritical fission chain reaction cannot result from any moderating conditions. Furthermore, the dose assessment for the case of complete loss of water revealed a maximum dose rate of 5.52 ± 0.045 mSv/h and 288 ± 4.65 𝛍Sv/h at the tank surface and tank platform, respectively. These values are significantly above background radiation levels; however, the calculated spatial distribution is highly asymmetric and the dose rate falls off rapidly with increasing distance. These results suggest that a facility-level emergency response is sufficient to address the accident. Our findings can augment the existing facility emergency procedures, thereby contributing to the improvement of radiological safety.