Operation of Tehran Research Reactor With LEU Fuels.
Mohammad Zaker
Atomic energy Organization of Iran.
Nuclear Research Center
Reactor Division
P.O.BOX 11365-8486
North Kargar Ave
TEHRAN - IRAN

Abstract
The 5 MW Tehran Research Reactor (TRR) has been converted low Enriched Uranium (LEU) fuel and since beginning of 1993 the reactor operating with LEU fuel. According fuel management program several core configuration should be created in order to reach to the equilibrium core. The transitory cores have different cycle length and totally it takes 516 full operation days to reach to the equilibrium core. Considering the TRR operation schedule we are approaching to the equilibrium core. This paper presents the possible cores configuration sequence that applied to TRR from first operating core to the equilibrium core.

Neutronics, Thermal Hydraulics and Safety Parameter Studies of a 3 Mw Triga Mark Ii Research Reactor.
M.D. Quamrul Huda
Institute of Nuclear Science & Technology,
Atomic Energy Research Establishment
Dhaka, Bangladesh

Abstract
This study deals with the analyses of neutronics, thermal-hydraulics and safety parameters of the current core configuration of a 3 MW TRIGA MARK II research reactor of Atomic Energy Research Establishment (AERE) at Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA reactor. A relatively small but complex and non-periodic case of research reactor geometry seems to be a good choice for benchmark calculations with Monte Carlo method. Continuous energy cross-section data from ENDF/B-VI and ENDF/B-V and S(,) scattering functions from the ENDF/B-IV library were used. Because of the unavailability of cross sections of ¹⁶⁶Er and ¹⁶⁷Er nuclides in the MCNP cross section library, NJOY99.0 nuclear data processing system was used to generate these two cross section files which are essential for TRIGA analysis. MCNP4C input was prepared in such a way that any desired core configuration could be simulated easily. The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics was established by benchmarking the TRIGA experiments. Most of the steady-state experiments were simulated; effective multiplication factors, control rod worths, excess reactivity and shutdown margin as well as the neutron flux and power distributions were used in the validation process of the physical model. The MCNP calculated values of the multiplication factor are consistent with the experimental data, although the MCNP calculated value underestimated 0.2510%k/k for control rods at critical positions and overestimated 0.0966%k/k for all control rods withdrawn positions. The reactivity curves calculated by MCNP are also consistent with the experiments within the estimated error of 10%. Calculations of fast neutron flux, and fuel and graphite element worths distribution are also presented. Good agreement between the experiments and MCNP calculations indicate that the simulation of TRIGA reactor is treated adequately.

Thermal-hydraulic analyses of the reactor operating under both steady-state and transient conditions were performed by PARET thermal-hydraulic code. Results of the neutronic analysis performed by MCNP were used in PARET input to study the thermal hydraulic behavior of the reactor to predict the safety parameters. To benchmark the PARET model, data were obtained from different measurements performed by thermocouples in the Instrumented Fuel (IF) rod during the steady-state operation both under forced and natural convection mode and compared with the calculation. The mass flow rates needed for input to PARET were taken from the FSAR for a downward forced coolant flow equivalent to 3500 gpm. For natural convection cooling of reactor, mass flow rate was generated using NCTRIGA code. The power output of the reactor limited by three dependent thermal and hydrodynamic variables: the DNB ratio, the maximum fuel temperature (T_max), and the core pressure drop (p) are discussed in this study. At a design flow rate of 3208.9 kg/m².s (3500 gpm) with a hot rod factor of 1.854 the DNB ratio is found to be 2.7851 and the pressure drop is 1.02 kPa along the hot channel at a power level of 3 MW. The peak heat flux occurs at the axial center of the fuel elements; therefore the DNB ratio is minimum at this location. The maximum power density remains, by a substantial margin, below the level at which DNB could occur. Peak fuel temperatures measured by thermocouples in the IF rods at different power levels of the TRIGA core were compared with the values calculated by PARET. Fuel surface heat flux, heat transfer coefficients, and axial distribution of temperatures of the fuel centerline, fuel surface and the cladding `surface in the hot channel were calculated for the reactor operating at full power level. The investigated results were found to be in good agreement with the experimental and operational values.

The testing of the PARET model calculations through benchmarking the available TRIGA experimental and operational data for pulse mode operations showed that PARET can successfully be used to analyze the transient behavior of the reactor. Major transient parameters, such as, peak power and prompt energy released after pulse, FWHM of pulse peak, maximum fuel centerline temperatures for different fuel elements at different pulses etc. were computed, and compared with the experimental and operational values. It was observed that pulsing of the reactor inserting an excess reactivity of $1.996 shoots the reactor power level to 873 MW compared to an experimental value of 852MW; the maximum fuel temperature corresponding to this peak power was found to be 512^oC. The investigation on maximum available reactivity insertion at full power ($2.24) by the transient rod raises the reactor power to 1629 MW and the fuel centerline temperature from calculations is found to be 937^oC.

Loss-of-flow accident analysis conducted on the forcedflowcooled TRIGA core shows that the flow reverses direction to the natural convection mode very quickly and smoothly, with essentially no interruption in the fuel temperature decay rate. Thus, the afterheat from the shutdown reactor will be removed by natural convection following pump failure or emergency shutdown. The loss-of-flow transient after a trip time of 4.08 sec at 85% of loss of normal flow for the TRIGA core shows a peak temperature of 709.22^oC in the fuel centerline and 131.94^oC in the clad and 46.63^oC in the coolant exit of the hottest channel. The transient was terminated at 15% of nominal flow after about 48.0 sec. Realistically, the flow would be expected to reverse direction and establish a natural convection flow rate that should be adequate to cool the core. The time at which the reversal of coolant flow starts is about 67.0 sec.

Back to Top

The History of the SLOWPOKE2 in Jamaica.
Gerald Lalor.
International Centre for Environmental and Nuclear Sciences.

Abstract
In the early 70's the view that scientific programmes on Jamaica were valuable for socio-economic development was growing. After some years of effort funding was obtained mainly from the EEC  and GOJ for a small multidisciplinary centre built around a SLOWPOKE reactor to help further this development.  The main thrust of ICENS at this time is on environmental geochemistry and its applications to agriculture and health.

The successes include the elemental mapping Jamaican soils and waters, the mitigation of lead poisoning among children, and the relationships between the elemental compositions of food and soil. The potential health effects of heavy metal ingestion are growing in importance. ICENS welcomes collaborations in its programmes.

Nuclear Analytical Methods at ICENS Part 1: Operations and Radiation Safety Aspects of the SLOWPOKE-II Reactor In Jamaica.
Charles Grant, G. Lalor, J. Preston and M. Vutchkov
International Centre for Environmental and Nuclear Sciences
University of the West Indies,
Mona, Jamaica

Abstract
The production of argon 41 by the activation of air in the irradiation tubes of the SLOWPOKE-II research reactor was investigated for various irradiation and counting conditions. Gamma spectroscopy was used to measure argon 41 in the lab and the data used to develop a simple model to estimate exposure to the analyst based upon reactor power and irradiation time. The data obtained from the simple model were then compared with the corresponding Derived Air Concentration limits. The results show that, with the current irradiation schemes and work conditions that the radiological consequences from the immersion in the semi-infinite radioactive Ar-41 cloud produce from the activation of air are minor.

SLOWPOKE: Early History and Development.
J. Hilborn

Abstract
The SLOWPOKE research reactor was conceived in 1967 at the Whiteshell Laboratory of Atomic Energy of Canada Ltd. (AECL).  In 1970 a prototype unit was designed and built at Chalk River. It was primarily intended for Canadian universities, providing a higher neutron flux than available from small commercial accelerators, while avoiding the complexity and high operating costs of existing nuclear reactors. The Chalk River prototype was started up in 1970 and moved to the University of Toronto in 1971. It had one sample site in the beryllium reflector and operated at a power level of 5 kW. In 1973 the power was increased to 20 kW and the period of unattended operation increased from 4 hours to 18 hours. At 20 kW, the thermal neutron flux at the sample site was 10(12) n/cm².s. The first commercial SLOWPOKE was started up in 1971 at AECL's Commercial Products Division in Ottawa; and in 1976 a commercial unit was installed at the University of Toronto, replacing the original prototype unit. The commercial unit has five sample sites in the beryllium reflector and five sites in the water outside the reflector. Subsequently, five more commercial SLOWPOKES were installed in Canada and one in Jamaica. Six of the original reactors are still in operation and one has been refuelled. Although all of the technical goals were achieved, the lack of foreign sales was disappointing.

Back to Top