Patent Number: 047643390
Section: description

DETAILED DESCRIPTION OF THE INVENTION Reference will now be made in detail to the present embodiments of the invention, an example of which is illustrated in the accompanying drawings. Referring to FIG. 1, reactor 10 is comprised of a core which consists of two symetric segments 14 and 16. These segments are housed within a pressure vessel 12. Each of the fuel elements 14 and 16 are comprised of concentric circumferential fuel plates 13. The reactor coolant enters the pressure vessel 12 through coolant inlet 18 and follows the flow indicated by arrow 31. Some of the coolant passes through core segment 14 through coolant channels 11, which are formed between the concentric fuel plates 13. A portion of the inlet coolant flow is diverted such that it does not pass through core segment 14. In a preferred embodiment, the coolant bypasses core segment 14 through a channel which is formed between the outermost circumferential fuel plate and the inner wall of pressure vessel 12. The outlet coolant from the first core segment 14 is mixed with the bypass flow in a coolant mixing plenum 28, which is formed between core segments 14 and 16. The coolant flow and mixing in plenum 28 is indicated by arrows 32. The coolant then flows through the coolant channels in core segment 16 and through the coolant outlet 20 as indicated by arrow 33. Preferably, core segments 14 and 16 are supported by core support column 26 which is integrally attached to the ends of pressure vessel 12. In a preferred arrangement, core support column 26 is hollow such that reactor control rods 35 may be inserted and retracted as needed. Fuel core segments 14 and 16 are designed with thin plates 13 and narrow coolant channels 11 to maximize the fuel plate surface area per unit core volume so that core cooling capacity is maximized for high core power densities. The segmented core arrangement of reactor 10 has several advantages when compared to conventional high flux reactors. The coolant passing through the individual core segments is exposed to only a very short heated flow path so that the coolant temperature rise is relatively small. The outlet coolant from the first core segment 14 is mixed in the central mixing plenum 28 with bypass inlet coolant that has passed between upper core segment 14 and the inner wall of pressure vessel 12. This mixes the hot-stripe outlet coolant and results in a low inlet temperature to lower core segment 16. Thus, the critical value of the peak coolant outlet temperature can be kept relatively low in such a split core arrangement. Furthermore, turbulent heat transfer effects at the entrance to lower core segment 16 enhances cooling of the peak power density near the core midplane. The net result is that the split core configuration can operate at substantially lower peak coolant temperatures than its conventional counterparts, even at modest coolant flow rate and pressures. Additionally, the resultant increase in critical heat flux safety margin and the lower fuel plate temperature (and lower oxide build-up rate) allows for operation of the double core configuration at sufficiently high core power densities to attain the required 10.sup.16 n/cm.sup.2 s. Core segments 14 and 16 may be comprised of a plurality of pie-shaped segments 17, to facilitate fabrication and accessability. It will be readily apparent to those skilled in the art that other means may be used as a coolant bypass channel. In a preferred embodiment of the present invention, central core support column 26 will have aperatures 27 in the coolant mixing plenum region 28 and similar aperatures in coolant inlet region 29, thereby providing a means for bypassing first core segment 14. In other preferred arrangements in the present invention, core segments 14 and 16 are comprised of involute plates or spirula type rolled plates of fuel material as illustrated in FIG. 2. The plates 42 bend circumferentially around the core and are supported by a series of pads or pins 44. Spirula type plates 42 may be supported from pressure vessel 12 via support column 46, in an arrangement similar to the embodiment of FIG. 1. For the goal flux levels of the present invention, a core power density of 10 MW/L is required. It is therefore important to minimize the critical reactor volume in order to minimize the total reactor power. Therefore, preferrably the core fuel material is comprised of a high reactivity worth fuel material like fully enriched uranium or high density fuels like U.sub.3 Si.sub.2 or UAl.sub.x (uranium-aluminum alloys). Neutron beam access channels 22 are positioned outside of pressure vessel 12 in a pool of moderating material, represented by shaded area 24. The principal moderator of choice for the present invention is D.sub.2 O because of its longer neutron diffusion length and lower parasitic neutron absorption cross section. The use of a moderator with this combination of parameters places the region of peak thermal neutron flux several centimeters from the core interface. The peak thermal neutron flux region thus covers a large volume while maintaining reduced fast-neutron and gamma-ray contamination compared with an equivalent H.sub.2 O-moderated reactor. Since beam access channels 22 may be aimed at central mixing plenum region 28, the split core arrangement allows for direct radial beam access to the resulting high flux environment, in addition to tangential access, without exposing the direct field-of-view of the beam channels to significant high-energy-neutron and gamma contamination from the core. The reactor is operated on a hard neutron energy spectrum (fissions occur in the fast energy group with little moderation inside the fuel segments) to enhance the neutron leakage into the radial reflector region 24, where the neutron leakage can be moderated and accessed by the experiments. There is some thermalization and peaking in plenum region 28 but because of the long neutron mean free path in D.sub.2 O, the core segments are neutronically tightly coupled. The major thermal neutron flux peaking occurs in a large (more than 100 liter) torodial ring surrounding the core in the radial reflector pool 24. In the preferred arrangements of the present invention, pressure vessel 12 is placed directly adjacent to the core segments 14 and 16 for several reasons. First, it allows the heated core coolant to be isolated from the reflector pool 24, so that the beam tubes 22 and the hot and cold sources can be operated in a low temperature, low pressure environment. This low temperature, low pressure environment improves both safety and the experimental neutron economy. The experimental instruments can have thin windows and thin walls that are not part of the primary reactor coolant pressure boundary. Research instrument maintenance and modifications are also facilitated by the low temperature and low pressure environment. The reactor pressure vessel 12 itself is configured much like a small diameter pipe with flanges at both ends. This is a simpler, less expensive vessel to design and fabricate (in spite of the fact that the vessel lifetime is shorter because of the higher fast neutron fluence close to the core) than conventional reactor-pool pressure vessels with numerous beam tube penetrations. Second, the neutron energy spectrum directly adjacent to the core is quite hard (high energy), so that the pressure vessel is relatively transparent to the core leakage flux. Third, it places less material between the core and the experimental instruments because the beam tube does not have to be designed to withstand a high external pressure. For the above reasons, it is, therefore, advantageous to place the pressure boundary here rather than around the beam tubes and cold sources where it would result in a large parasitic loss of thermal neutrons and increased gamma background. Typical design characteristics for an ultra-high flux reactor with two core segments are listed in Table II. Although the present invention has been described with reference to a reactor having two core segments, it will be readily apparent to those skilled in the art that the present invention may also comprise more core segments. TABLE II ______________________________________ DESIGN CHARACTERISTICS OF AN ULTRAHIGH FLUX DOUBLE SEGMENT REACTOR ______________________________________ Core Dimensions Fuel donut height (cm) 13.7 Fuel donut ID (cm) 20.0 Fuel donut OD cm) 40.0 Fuel donut thickness (cm) 10.0 Core volume (liters) 25.8 Central plenum height (cm) 10.0 D.sub.2 O bypass gap (cm) 2.0 Zr gamma shield thickness (cm) 0.5 Zr pressure vessel thickness (cm) 1.3 Fuel Material U.sub.3 Si.sub.2 /Al Uranium enrichment (w/o U-235) 93 Volume fraction U in fuel seat 0.45 Volume fraction void in fuel meat 0.1 Maximum fuel density (g U/cc meat) 4.8 Fuel plate thickness (cm) 0.102 Coolant channel thickness (cm) 0.076 Fuel meat thickness (cm) 0.051 Cladding material Aluminum-2219 Cladding thickness (cm) 0.025 Number of fuel plates/assembly 56 Side plates material Aluminum-2219 Side plate thickness (cm) 0.5 Fuel burnup limit (fission/cc) 2 .times. 10.sup.21 Physical Characteristics Reactor power (Mw) 300 at BOC, 275 at EOC Cycle length (days) 14.0 Core average power density (MW/L) 11.6 at BOC Power peaking factor (peak/average) 1.6 at BOC Peak power density (MW/l) 18.5 at BOC Peak reflector thermal neutron flux, 1.0 .times. 10.sup.11 E &lt; .683 eV (n/(cm.sup.2 s)) Gamma environment in region of peak 10 neutron flux (W/g in D.sub.2 O) Fast flux (E &gt; 1 MeV) contamination 1.1 .times. 10.sup.14 in region of peak thermal flux (n/(cm.sup.2 s)) Core fissile loading at BOC (kg U-235) 22.2 Number of unique fuel loading zones 7 Core burnable poison loading at BOC 8.2 (g b.sup.14) Fuel burnup (kg U-235) 5.6 Thermal Hydraulic Conditions Coolant inlet pressure (MPa) 4.2 Core outlet pressure (MPa) 3.4 Pressure vessel design pressure (MPa) 5.5 Coolant flow rate (kg/s) in core channels 602 in bypass gap 820 Coolant velocity (m/s) in core channels 16.0 in bypass gap 27.2 Coolant inlet temperature (.degree.C.) 38 Coolant outlet temperature (.degree.C.) 108 hot channel Coolant T (.degree.C.), hot channel 70 Peak surface heat flux (MW/m.sup.2) 17.3 Hot spot fuel plate temperature, 234/387 BOC/EOC (.degree.C.) Margin to CHF (std. dev.) 3.3 Margin to hydraulic instability 6.8 (standard deviation) Oxide thickness at EOC (mm) 0.046 Margin to fuel melting (.degree.C.) 262 ______________________________________ The present invention thus provides a nuclear reactor capable of producing a neutron flux an order of magnitude larger than that produced by present day high flux reactors. Flow instability effects are alleviated in the present invention by the short flow channels and the hot-stripe mixing in the coolant plenum between the core sections. The core, which is comprised of thin circumferential fuel plates, produces a larger fuel plate surface area per unit core volume, shortened flow paths, and efficient hydraulic geometries. The life-limiting aluminum oxide layer formation on the fuel cladding, which insulates the plates and drastically increases fuel temperatures is not considered to be a key constraint of the split core concept. This results from the significantly lower peak operating temperatures acheivable with the short-flow-path and increased plate surface area of this invention. The present invention also enhances accessability to the high flux environment with radial beam tubes which are disposed in a low temperature, low pressure environment. The foregoing description of the preferred embodiments of the invention have been presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise form disclosed, and obviously many modifications and variations are possible in light of the above teachings. The embodiments were choosen and described in order to better explain the principle of the invention and its practicable applications to thereby enable others skilled in the art to best utilize the invention and various embodiments and with other modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto.