9X9 Lattice - Reduced Average LHGR
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NFI 9x9 High Burnup BWR Fuel
- Proven Design Hydraulic Mechanism for High Precision Control of Coolant Leakage between LTP and Channel Box
The spacer, upper tie-plate (UTP) and lower tie-plate (LTP) play important roles for the thermal-hydraulic performance. The ring-type spacer provides a high critical power which has been demonstrated by testing , The low pressure drop spacer and UTP compensate for the relatively high frictional pressure drop of the 9 x 9 FA. With respect to the pressure drop, the advanced 9 x 9 FA has been proven by thermal-hydraulic testing including the channel stability test . The LTP design has a unique feature of the hydraulic mechanism for sealing the gap between the LTP casting and the channel box. The flexible LFCD (Leakage Flow Control Device) made of inconel sheet effectively seals the leakage flow which otherwise increases with burnup due to the channel-box bulge, thereby providing high precision control of coolant leakage to the core bypass. The performance of the LFCD has been proven by hydraulic testing and NFl's operating fuel experience has shown its effectiveness.
Further development of the spacer design (ULTRAFLOW) has been made and provides 8% higher critical power and 2% lower pressure drop than that of the ring type spacer, when it is used for the 9 x 9 FA , Either operational margin to the minimum critical power ratio (MCPR) or stability of the core decreases with the more negative void coefficient of the MOX fuel assemblies. The higher critical power provided by the new spacer accommodates the larger margin of MCPR. The lower pressure drop of the spacer increases the channel stability, which in turn increases the core stability.
All nuclear designs to be discussed assume a Pu material from the reprocessing of the BWR spent fuel with burnup of 39.5 GWd/t. Total fissile Pu content becomes 63.4 wt% with decay of Pu241 into Am-241 for 8 years after the discharge. The depleted uranium (0.2wt% U-235) is used for the MOX carrying material. A standard fuel active length of the U02 FA is about 3.7 m, but for the MOX fuel in this study a short active length of 3.55 m with extended rod plenum was used to assure sufficient margin against potentially higher FGR from the MOX fuel. The burnable absorber (gadolinia) was used in the U02 fuel rods, but not in the MOX fuel rods.
The enrichment designs that will be discussed below are shown in Fig. 2 and Fig. 3 for the average discharge burnup of 39.5 GWd/t and 45 GWd/t, respectively. Both designs use only 2 enrichment Pu levels. The four corner rods and the gadolinia bearing rods contain uranium but not Pu. The enrichment of the Pu or uranium as well as the gadolinia content are axially uniform over the active length. The current MOX fuel design practice for both PWR and BWR is to use 3 to 6 Pu enrichment levels in order to decrease the higher local peaking factor , Such design may be best utilized to increase the MOX inventory, but the fabrication becomes complex. The total cost of the MOX utilization in the LWR is dependent on the capital and running costs of the MOX fuel fabrication plant. Simplified BWR MOX design as proposed in this study is a compromise of these factors, but the MOX inventory loss (corresponding to 4 corner rods) is only 6 %. Another important design merit pertaining to the LSWC is relatively lower maximum Pu enrichment. The maximum fissile content (Puf) is only 3.9 wt% or 4.5 wt% for the high burnup design of average discharge burnup of 39.5GWd/t and 45GWd/t, respectively. This is in sharp contrast with the other type BWR FA, e.g. 9 x 9FA with one small water rod . In the MOX FA fabrication plant design, the lower the maximum enrichment, the larger the batch size in Pu handling which may be limited by the criticality control and the shielding design. When the batch size becomes large, availability of the equipment is increased, thus the MOX fabrication cost is reduced.
2.3 CORE PERFORMANCE OF THE ADVANCED 9x9 BWR MOX FUEL
A reference plant in this study is a 1100 MWe class BWR with 764 fuel assemblies in the core. An equilibrium core of the cycle length of 13 effective-full-power-months (EFPM) was investigated for the core performance of the advanced 9x9 BWR MOX fuel. Table I shows the results of the core performance calculations. In this Table, Case U45 represents a reference core of the U02 fuel assembly, Case M39 and Case M45 for the uniform core of the MOX fuel assemblies, and Case M39/U45 and Case M45/U45 for the mixed core. The figures "45" and "39" after the letter "U" or "M" mean the average discharge burnup of 45 GWd/t and 39.5 GWd/t, respectively. It is obvious from these calculation results related to MLHGR, MCPR and Shutdown Margin that the steady-state core performance of the MOX FA
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