Pressure Tube Heavy Water Reactors Research Articles

High-Utilization Lattices for Thorium-Based Fuels in Heavy Water Reactors

New fuel bundle and lattice concepts to implement thorium-based fuel cycles in pressure tube heavy water reactors (PT-HWRs) have been explored to achieve maximum resource utilization. As an existing, operational technology, PT-HWRs are highly advantageous for implementing the use of thorium-based fuel cycles because of their high neutron economy and online refueling capability. A PT-HWR is flexible in that it can use one, two, or more different types of fuels in either homogeneous or heterogeneous cores to optimize power production, fuel burnup, and new fissile fuel production. In a heterogeneous PT-HWR core, higher fissile content seed fuel will be optimized for power and excess neutron production, and lower fissile content blanket fuel will be optimized for production of 233U. Five different lattice concepts were investigated for potential use in a once-through thorium cycle in a PT-HWR. The lattices involved 43-, 35-, and 21-element bundles with a central cluster of ThO2 pins, or a Zircaloy-4 (Zr-4) central displacer tube containing either stagnant D2O coolant or solid ZrO2, to help reduce coolant void reactivity (CVR). The fuel in the outer pins is a homogeneous mixture of Th and low-enriched uranium (LEU) (~5 wt% 235U/U) or reactor-grade Pu (~67 wt% fissile). The content of the LEU or Pu was varied to achieve different levels of burnup, and it is presumed that low-reactivity fuel would be used as blanket bundles. It was found that the various lattice concepts could achieve burnups ranging from ~10 to 80 MWd/kg and that the fissile utilization could be up to 60% to 100% higher than what is currently achieved in a PT-HWR using natural uranium fuel. Burnup-averaged CVR ranges from approximately +1 to +16 mk (1 mk = 100 pcm = 0.001Δk/k), depending on lattice type and fuel composition. Assuming a maximum linear element rating of ~50 kW/m, the maximum permissible bundle power ranges from ~520 to 800 kW.

Axial Dependence of Partial Void Reactivity in a Light Water-Cooled, Heavy Water-Moderated, Pressure-Tube Reactor

The effect on reactivity of changes in the coolant levels in the pressure tubes of a pressure-tube heavy water reactor is experimentally studied to clarify the effect of an axial coolant void fraction distribution. The coolant void fraction distribution is simulated by stepwise changes in the coolant levels in the Deuterium Critical Assembly (DCA). The reactivity is measured for a 25.0-cm-pitch square-lattice core with a positive coolant void reactivity. The reactivity changes resulting from changes in the coolant levels are measured as changes in the critical heavy water levels. The axial distribution of the thermal neutron flux is also measured by the copper activation method.In these measurements, the critical heavy water levels show a sinuous curve having a maximum and a minimum, and a positive reactivity larger than those of both the 0% and the 100% void uniform cores is introduced at certain coolant levels by stepwise changes in the coolant levels.An experimental analysis is performed with the coupled WIMS-ATR/CITATION code system, whose analytical method was established through DCA critical experiments. Agreement between experiment and analysis is fairly good. Furthermore, the peculiar reactivity behavior resulting from changes in the coolant levels is analyzed using a simplified model to take note of typical reactor physics parameters. It is clarified that this anomalous phenomenon is caused by the combined effect of the flattened S curve change in the thermal neutron absorption and the even flatter S curve change in the neutron leakage caused by the changes in the coolant levels. Useful information is obtained regarding reactivity behavior with an axial coolant void fraction distribution.

General Empirical Model for60Co Generation in Pressurized Water Reactors with Continuous Refueling

A simplified model is presented that permits one to calculate the average activity on the fuel elements of a reactor that operates under continuous refueling, based on the assumption of crud interchange between fuel element surface and coolant in the form of particulate material only and using the crud specific activity as an empirical parameter determined in plant. The net activity flux from core to out-of-core components is then calculated in the form of parametric curves depending on crud specific activity and rate of particulate release from fuel surface. In pressure vessel reactors, contribution to out-ofcore radionuclide inventory arising in the release of activated materials from core components must be taken into account. The contribution from in situ activation of core components is calculated from the rates of release and the specific activities corresponding to the exposed surface of the component (calculated in a straightforward way on the basis of core geometry and neutron fluxes). The rates of release can be taken from the literature, or in the case of cobalt-rich alloys, can be calculated from experimentally determined cobalt contents of structural components and crud. For pressure vessel reactors operating under continuous refueling, activation of deposited crud and release ofmore » activated materials are compared; the latter, in certain cases, may represent a sizable (and even the largest) fraction of the total cobalt activity. It is proposed that the ratio of activities of /sup 59/Fe to /sup 54/Mn may be used as a diagnostic tool for in situ activation of structural materials; available data indicate ratios close to unity for pressure tube heavy water reactors (no in situ activation) and ratios around 4 to 10 for pressure vessel, heavy water reactors.« less