Neutronic Analysis of LEU-started Molten Chloride Fast Reactor without Fuel Reprocessing

One of the rarely explored molten salt reactor (MSR) designs is the molten chloride fast reactor (MCFR). This MSR design employs chloride salt instead of fluoride and operated in a fast spectrum. MCFR brings all the advantages of an MSR including breeding whilst being able to burn plutonium and minor actinides efficiently. Since not many countries have access to civilian plutonium, MCFR can also be started using low-enriched uranium (LEU). This study is an initial neutronic analysis of an MCFR using LEU as its startup fuel. Parameters analyzed are conversion ratio (CR) and its neutronic safety, namely effective delayed neutron fraction (βeff), temperature coefficient of reactivity (TCR), and void coefficient of reactivity (VCR). The core is divided into Core Zone and Blanket Zone. The fuel composition of NaCl-UCl3 with a molar fraction ratio of 60:40 and 50:50 is used in Core Zone and Blanket Zone, respectively. The neutronic calculation is performed using MCNP6 code with ENDF/B-VII library. For reference geometry, CR is valued at 0.9298, βeff at 0.00731, TCR at -19.8 pcm/°C, and average VCR at -154.31 pcm/void%. Thereby, the MCFR fulfills inherent safety criteria. Although its value is remarkably high, CR can be further optimized by modifying the separator and reflector material.

INVESTIGATION ON INHERENT SAFETY OF ONE FLUID-MOLTEN SALT REACTOR (OF-MSR) WITH VARIOUS STARTING FUEL

Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

Experimental Estimation of Moderator Temperature Coefficient of the IPEN-MB-01 Reactor

The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28x26 rectangular array of UO2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4oC, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and (Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them.

STUDY ON TEMPERATURECOEFFICIENT OF REACTIVITY FOR PEBBLE BED REACTORWITH THORIUM FUEL

The pebble bed reactor has a flexibility to utilize a wide range of different fuel cycles without significantly modifying the reactor core geometry. These fuel cycles can be comprised of various fissile and fertile fuel isotopes. The purpose of this paper is to study the temperature coefficient of reactivity for pebble bed reactor fuelled by thorium. The HTR-Modul was chosen as the reactor model. A series of calculations was performed by the use of the Monte Carlo transport code MCNP6 with the continuous energy nuclear data library ENDF/B-VII. The k∞ calculation results show that the fuel load with a mass of 2g Th/pebble, 5g Th/pebble and 20g Th/pebble can achieve a critical core with discharge burnup of 65,644; 101,500; and 114,215 MWd/t, respectively. The calculation results of temperature coefficient conclude that specific mass of thorium per pebble might have a positive temperature coefficient of reactivity at a certain temperature. Further investigation needs to be conducted to analyze the behavior of these temperature reactivity coefficients in more detail.