Fuel lattice

Early design calculations indicated that the following conditions could be expected in the reflector and fuel lattice ... 

In the first group of experiments the northwest thimble was left open. Fluxes were measured- in the fuel lattice., the, northwest quadrant of the graphite reflector, and at other uiscellaneous points in the structure. 

Reactivity is controlled by rods consisting of articulated absorber elements formed from hollow cylindrical sections of boron carbide (65 mm diameter x7-5 mm thick) sheathed in the annulus between two aluminium alloy tubes of 70 mmx2 mm and 50 mmx2 mm respectively. They are inserted or removed from the core at a rate of 0-4 m/s (the 12 local automatic control rods are withdrawn at 0-2 m/s) by individual servomotors installed at the top of the control rod channels. With the exception of the automatic rods, all the rods are fitted with graphite followers so that, as they are withdrawn, they are not replaced by water. The square lattice of 211 control rods and 12 vertical power profile sensors has a pitch of 700 mm and is angled at 45 to the fuel lattice. The channels are made... 

Moderate power density, Pb coolant instead of combustible Na with its small boihng margin, a widely spaced fuel lattice instead of the close-packed one chosen for the moderating Na, Pb temperature gain reduction, and increase in its temperature at the core inlet upward of 400°C with a margin to Pb melting, no fuel assembly shrouds (as in PWR with a widely spaced lattice)... 

With a wider-spaced fuel lattice in the LFR it is possible to increase the coolant flow area and to reduce its velocity. As a result, the hydraulic resistance and the pumping power requirements in the LFR are a factor of 3—5 smaller than in the SFR. The lead coolant temperature gain reduced down to 100-120°C allows raising its temperature at the core inlet to 400—420°C and providing a sufficient margin to its crystallization point (T<-fyst = 327°C). Even so, the fuel cladding temperature in the hot spot will not exceed 650° C, whereas in the sodium-cooled reactors it is higher than 700°C. 

Widely spaced fuel lattice with a large coolant flow area - for reducing its velocity to 2 m/s and the hydrauhc losses, to 0.10-0.15 MPa, as well as for increasing the amount of heat removed by natural circulation. 

The core consists of 151 hexagonal fuel assemblies (FAs) (size across flats is 234 mm) with fuel elements and fuel lattice parameters analogous to those in VVER-1000. Each FA contains boron carbide rods which are combined in a cluster to form a control device. The control devices of 135 FAs are connected to drives of the electromechanical control and protection system (CPS). The core height is 3.53m, its equivalent diameter is 3.05m at average power density of 69.4 kW/1. 

Lattice Parameter Measurements for o Concentric Tube Fuel Element, D. E. Wood, K. R. Bimey, and E. Z. Block (GE-HAPO). Lattice parameters have been measured for a concentric tube, natural uranium fueled lattice, moderated by graphite. The experiments were made to test calculatlonal models for lattices with toe complex fuel geometry involved. For toe 10-1/2-in. lattice, k >, t, p, and < were measured with water and air In toe co(dant diannels. For toe 8-3/8-in. lattice, k and f were measured with water coolant only. 

Fuel Lattice Pilch (in.) Measured Bucklii (cm ) X 10 CMcttlaied... 

Two plutonium compositions were used, as listed in Fig. 1, vdiich also shows the designs of the coextruded aluminum fuel assemblies. Hie experimental lattices consisted of 19 of these fuel assemblies in hexagonal arrays at 12.12- or 14.00-in. pitches. To provide an easily calculated radial boundary, the fuel lattices were extended for two additional rings with nonfissioning iron-lithium assemblies having very nearly the same absorption cross section as the fuel. 

Organic-Cooled Heavy - Water - Moderated U-233-Fueled Lattice Experiments. G. A. 

G. A. PRICE et al., Organic-Cooled, Heavy Water-Moderated, Fueled Lattice Ejqieriments, BNL 50012, (T-434), Brookhaven National Laboratory (Aug. 25, 1966). 

Critical Experiments on Cluster-Type Fuel Lattices of ATR in Japan, B. Sakata, Y. Hactdya, K. Shiba, N. Fukutmra, A. Nishi (PRNFDC-Japan)... 

The material buckling, Bm, was Obtained from axial and radial neutron flux distribulions by Cu wire activation measurements in the case of 1.2% enriched UO fuel lattices. For the 0.7 and 1.5% UO fuel lattices which were nmde critical by being surrounded with the 1.2% fuel lattice region, Bm was measured by the progressive sub-stttution method. Fine distributions of thermal-neutron flux were also measured by Dy f(dl activation. 

Increase of void ratio means decrease of H2O coolant in the cluster fuel lattice. The slowing down effect by hydrogen decreases values of 0, p, and critical mass. This tendency is strongly observed in 22.Sfrcm pitch.lat> tice but weak in the 25.0 cm one. The maximum value of Bm Is seen at 30% void ratio, R is concluded that la the 25.0-cm pitch lattice, contribution of slowing doym effect by ikO coolant is surpassed by its thermal-neutron ab-sori>tion effect. 

KENO-IV. calculates toe subcritical reactivity for these massive, highly mulched U-A1 alloy fuel lattices moderated by H>0 over a wide range of reactivities with... 

The experiments were performed at B W s Lynchburg Research Center in the Company s CX-10 critical facility. CX-10 is a tank-type critical facility licensed for the performance of critical experiments with water-moderated UOj and mixed oxide (Th-U) fuel lattices. Although the facility is licensed to use either light water or heavy water as a moderator, only light water was used in this series of experiments. 

One phase cif. this research invoived critical experirhents with low-enriched uranium fuel rods with slabs of uranium or lead on two sides immersed in water. Rather interesting results were noted in this research. Bierman observed that a lead or a depleted uranium wall (0.2 wt%< U) backed by water was a better reflector than water alone. Moreover, for the case of uranium shielding walls there was an optimum water gap spacing between fuel and wall that resulted in a more efficient reflector combination than if a d< y fitting uranium wall were used. The lead wall did not show this effect but was most reactive when positioned at the fuel cell boundary. As a result of these findings, a calculational study was made to examine the effects oh criticality of changing various parameters. These induded distance fiom wall to fuel, lattice pitch, wall tiiickness, and fuel compoa-tion. The results of this study substantiate the experimental results found by Bierman, and further indicate tiiat the effect of the metal reflector on criticality varies with all of the above-mentioned parameters. 

A Critical Nixed Oxide (U02-Pu02) Fuel Lattice Moderated ... 

Approaches to remove and transport heat from the fuel lattice and drive energy... 

An optimized fuel lattice with improved neutron moderation permits fuel bum-up increase in the ABV reactor core, see Fig. V-3. 

The elimination of soluble boron control together with the adopted parameters of the fuel lattice provide negative reactivity coefficients on the fuel and coolant temperature negative steam and integral power coefficients of reactivity in the entire range of operating parameters, which altogether secures inherent safety features of the reactor core. These inherent safety features ensure power self-control in a steady state reactor operation, power rise self-limitation under positive reactivity insertions, self-control of the reactor power and primary coolant pressure and temperature self-limitation in transients, as well as the limitation of the heat-up rate in reactivity-initiated accidents. 

To provide maximum fuel burn-up and considering the accepted limitation on uranium enrichment, the fuel lattice parameters that affect water-uranium ratio have been optimized. 

Figure XXII-7 shows the SSTAR core map. The fuel lattice consists of cylindrical fuel rods arranged on a triangular pitch (the hexagonal geometry does not imply that the core is formed of individual hexagonal fuel assemblies or bundles it merely reflects the assumed nodalization used for neutronics modelling.) A central two low enrichment zones blanket, the three enrichment zones, and locations for shutdown and control rods are indicated in the figure.

The result (10.255) may also be used to obtain the criticality condition for the reactor with finite fuel lattice but with infinite reflector (composed of the same moderator material). It is necessary to impose the latter condition because otherwise expression (10.243) for the slowing-down density would no longer be valid and new functions satisfying boundary conditions at the surface of the reflector would be required. The appropriate neutron-balance relation in terms of the flux at the various fuel-rod locations is then given by... 

A natural circulation system that was not considered in the main text is a pool type system. This would employ the same fuel lattice chosen for the natural or forced circulation systems evaluated in Appendix E and may involve some tradeoff of advantages and disadvantages. 

The R-Z two-dimensional core calculation model, as described by Fig. 2.30, may be a good first approximatiOTi to calculate a fast reactor core with a relatively simple loading pattern of hexagonal fuel assemblies (a tight fuel lattice). In such a configuration, the spatial dependence of the fast neutron flux is small and the rough estimation by the R-Z two-dimensional model may be applicable. 

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