Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program
Abstract
In order to significantly increase the power density of Light Water Reactors (LWRs), the
helical-cruciform (HC) fuel rod assembly has been proposed as an alternative to traditional
fuel geometry. The HC assembly is a self-supporting nuclear fuel configuration consisting
of 4-finned, axially-twisted fuel rods closely packed against one another in a square array.
Within the LWR core, HC fuel would in theory possess several inherent advantages over
traditional fuel, potentially allowing for operation at a higher power density. Chief among
these advantages are a larger surface-to-volume ratio, a shorter radial heat conduction path,
and improved mixing characteristics.
In previous work, computational models of the HC fuel assembly have been of limited
accuracy due to the absence suitable correlations. To address needs within these subchannel
analysis models, experimental measurements of rod bundle coolant mixing have been
conducted with 4x4 arrays of HC test rods. The tests used the technique of a hot water
tracer injection (at 95°C) into a bulk flow of cold water (at 25°C). Downstream temperature
measurements were used to judge the rate of lateral cross-flow within the HC rod bundle.
These tests were conducted at atmospheric pressure, and encompassed a range of mass
fluxes from 1000 kg/m2s to 3500 kg/m2s, HC rod twist pitches of 200cm, 100cm, and
50cm, and different hot water injection velocities and mixing lengths.
Data from over 300 tests was analyzed, yielding a best fit correlation for use with any twist
pitch, rod length, or coolant flow rate. Compared to the bare rod bundle, this correlation
implies an enhancement in the intensity of turbulent interchange of 40% brought about by
the HC geometry, and a 1.6% forced diversion of axial flow per subchannel, per quarterturn
along the rod length. These parameters fit all data points considered within a standard
deviation of 24%. Stochastic error was limited to ±16% by the use of precise temperature
sensors.
By applying this empirical mixing model to the subchannel representation of a BWR core
featuring the HC rod design, a need to increase the flow area of the edge subchannels was
demonstrated. This prompted a slight re-design of the HC fuel rod cross-section in order to
make room for small spacer protrusions at the duct wall, to increase flow to peripheral
subchannels. The modification was accomplished by reducing fin length, but increasing the
inner diameter to maintain the reference fuel volume. The water rod region was also
adjusted to maintain the reference assembly hydrogen to uranium atom ratio. With this
modification, the model predicted a 24% allowable power uprate for the 200cm twist pitch
HC core. Inlet and exit enthalpies were maintained from the reference cylindrical-rod core.
When applied to a PWR core of HC rods, also with a fixed power to flow ratio, this
empirical mixing model predicted an allowable power uprate of 47%, using traditional CHF
correlations for cylindrical fuel. In subcooled conditions, CHF is known to be more
sensitive to peaked areas of non-uniform heat-flux than in saturated two-phase flow
conditions. Therefore power density gains will likely be dependent on the degree to which
the rod twist would disrupt of nascent pockets of vapor; this effect should be further
investigated experimentally.
In order to further ascertain the potential gain in power density for the new design, an
experiment must be carried out to obtain CHF data for the HC rod bundle. Two facilities
with this aim were designed in great detail for BWR conditions: the first would operate
using high pressure water at 7MPa, and the alternate would use a relatively low pressure
refrigerant at equivalent conditions. The appropriate scaling laws were applied, which
resulted in the choice of R134a as the simulant fluid. The R134a facility was found to be
possible to construct at a greatly reduced cost.Tokyo Electric Power Corporatio