The IPR-R1 Research Reactor is a TRIGA Mark I type reactor. It is a pool reactor, and the fuel elements are cooled by water natural convection. The heat removal capacity of this process is great ...enough for safety reasons at the current maximum 250
kW reactor power level. However, a heat removal system is provided for removing heat from the reactor pool water. Power monitoring of nuclear reactors is always done by means of neutronic instruments. In the IPR-R1 the power is measured by four nuclear channels. This work presents the results and methodology for monitoring the power of this reactor by three thermal processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. The other two methods consist in the steady-state energy balance of the primary and secondary reactor cooling loops.
A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the ...Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.
The purpose of this paper is to present some of the main reasons, objectives, planning, and recent advances of an SCK·CEN/ININ joint project, which deals with the design and application of ...modern/expert control and real-time simulation techniques for the safe operation of a TRIGA Mark III research nuclear reactor. This project has been proposed as the first of its kind under a general cooperation agreement between the Belgian Nuclear Research Centre (SCK·CEN) and the National Nuclear Research Institute (ININ) of Mexico.
Temperature fields within the pool of the JSI TRIGA MARK II research reactor were measured to collect data for validation of the thermal hydraulics computational model of the reactor tank. In this ...context temperature of the coolant was measured simultaneously at sixty different positions within the pool during two transients. The obtained data not only revealed local peculiarities of the cooling water dynamics inside the pool but also allowed for better estimation of the coolant bulk velocity within the pool.
Many TRIGA
R
reactors are used as the source of thermal neutrons for neutron radiography. The majority of these applications involve steady state operation of the reactor which provides sufficient ...fluence for radiographic plates or real-time imaging. Many real-time applications would benefit from substantially increased fluences in a single frame (1/30 sec) or in a few frames if the number is limited by motion of the NR target. A pulsed TRIGA reactor can be useful in such cases. The low repetition rate for the standard pulsed TRIGA reactor (∽6/hour) has seriously limited this application. owever, the modern control rod drives and related digital instrumentation have made it possible to produce 60MW pulses at the rate of 1 at least every 5 to 6 seconds (∽10/ min or at least 100 times more frequent). The increased fluence in each of such pulses provides the capability to use electronic imaging with substantially increased resolution and signal-to-noise in a single frame. As an example. a 250kw steady state TRIGA reactor with L/D= 100 can provide a NR fiuence of about 1.8 x 10
9
n/cm
2
in 3 minutes i.e., with 45000 kw sec). A single frame of an electronic imaging system receives a fiuence from the same reactor corresponding to about 8.3 kw-sec, A TRIGA reactor pulsed every 6 seconds to 60MW will give a fluence in a single pulse corresponding to about 1800kw sec. This will allow a single frame to receive the fluence corresponding to the integration of more than 200 frames from the referenced steady state operation but without the electronic noise that accompanies extended frame integration. At least two areas of improved neutron radiography can result from use of the proposed rapidly pulsed TRIGA reactor. Not only would there be significant improvements in spatial and temporal resolution for radioscopy of moving NR objects but real-time cold neutron radiography would become more easily achievable.
The results of activation studies of TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete were irradiated in the reactor to simulate neutron activation ...in the shielding concrete. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides were measured in the samples with a high-purity germanium detector. The most active long-lived radioactive nuclides in the ordinary concrete samples were found to be
60
Co and
152
Eu. In the barytes concrete samples, the most active long-lived radioactive nuclides were
60
Co,
133
Ba, and
152
Eu. Activation in the concrete was also calculated using the ORIGEN2 code and compared to experimental results. Simple radioactive nuclide generation and depletion calculation using one-group cross-section libraries provided together with the ORIGEN2 code did not give conservative results. Significant discrepancies were observed for some nuclides. For accurate long-lived radioactive nuclide generation in reactor shielding, material-specific cross-section libraries should be generated and verified by measurement.
Studies were carried out for the design of a thermal, an epithermal, and a fast neutron beam for neutron radiography (NRG) in the horizontal beam port or in the thermalizing column at the Musashi ...reactor (TRIGA-II, 100 kW). Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, and > 10 keV, respectively. To produce a good neutron beam in terms of intensity and quality, several materials (bismuth (Bi), silicon (Si) and alumina (Al
2O
3)) were examined as neutron filters. The MCNP (Monte Carlo Neutron Photon) computations showed that a thermal, an epithermal, and a fast neutron beam could be produced using single-crystal Si, Al
2O
3 and Bi filters, respectively. Such neutron beams, if built into the Musashi reactor, could increase extensive application of NRG in Japan.
•For the validation of the Fusion Fission Hybrid Systems (FFHS) concept, the coupling of a fusion device to a multiplying fission medium can be seen as one very specific case of the coupling of an ...intense neutron source to a fission system.•Among the validation experiments performed in the past, some experiments in the frame of the MUSE program and the experiments pre-TRADE can be considered as an initial database to be used in a step-by-step validation of the FFHS concept.•A preliminary evaluation has been carried out by considering the coupling of a mono-energetic external fusion source (DD, En = 2.5 MeV, or DT type, En = 14.1 MeV) with the TRIGA RC-1 reactor in subcritical configuration.•A successive phase of validation should be envisaged, where the actual coupling of a small size fusion device with a multiplying fission blanket could be carried out.•Preliminary analyses show that some subcritical TRIGA RC-1 configurations can be representative of the neutron flux behavior in FFHS subcritical blanket.
Fusion Fission Hybrid Systems (FFHS) could have a potential role in the management of fission reactor wastes, with (in principle) some advantage over the comparable Accelerator Driven systems, devoted to the same objective. The validation of the concept poses major challenges in the area of fusion “source” development. However, the physics of the multiplying and transmutation blanket has also to be carefully understood and validated. Simple experiments can be envisaged for that purpose. Some of these experiments have already been performed in a different research framework (i.e. validation of the ADS concepts) and could constitute a very first database. Moreover, a new series of similar experiments could be planned.
Oregon State University is home to a TRIGA® Mark II reactor. In October of 2008, the reactor began operating on low enriched uranium fuel. A model of the facility exists in MCNP, a Monte Carlo code ...that can be used for criticality calculations. Until now, a bias in the calculation of the neutron multiplication factor has been carried forward from outdated core models.
This work involves updating various aspects of the model, including the geometry of the facility as well as materials and their properties, in order to arrive at a more accurate representation of the facility as it is today. The individual effect that each change has on the results of MCNP calculations of the core is documented.
Following the updates to the model, the model can emulate records that describe the startup of the reactor in October of 2008. The results of these calculations can be compared to actual data in order to establish a foundation for benchmarking the model and characterizing the reactor core. The deviation between calculated and expected results can be used to determine a single reactivity bias in the model.
The bias determined as a result of this work can be applied to future calculations using the model developed as a part of this work.
Graduation date: 2013
