STUDY OF ADVANCED LWR CORES FOR EFFECTIVE USE OF PLUTONIUM AND MOX PHYSICS EXPERIMENTS T. YAMAMOTO, H. MATSU-URA*, M. UEJI, H. OTA XA9953266 Nuclear Power Engineering Corporation, Toranomon, Minato-ku, Tokyo T. KANAGAWA Mitsubishi Heavy Industries, Ltd, Nishi-ku, Yokohama K. SAKURADA Toshiba Corporation, Isogo-ku, Yokohama H.MARUYAMA Hitachi, Ltd, Hitachi-shi Japan Abstract Advanced technologies of full MOX cores have been studied to obtain higher Pu consumption based on the advanced light water reactors (APWRs and ABWRs). For this aim, basic core designs of high moderation lattice (H/HM ~5) have been studied with reduced fuel diameters in fuel assemblies for APWRs and those of high moderation lattice (H/HM ~6) with addition of extra water rods in fuel assemblies for ABWRs. The analysis of equilibrium cores shows that nuclear and thermal hydraulic parameters satisfy the design criteria and the Pu consumption rate increases about 20 %. An experimental program has been carried out to obtain the core parameters of high moderation MOX cores in the EOLE critical facility at the Cadarache Centre as a joint study of NUPEC, CEA and CEA's industrial partners. The experiments include a uranium homogeneous core, two MOX homogeneous cores of different moderation and a PWR assembly mock up core of MOX fuel with high moderation. The program was started from 1996 and will be completed in 2000. 1. INTRODUCTION In Japan, the MOX fuel demonstration programs have been performed successfully and reload size use of MOX fuel has been prepared. The recycling of plutonium in light water reactors is expected to continue in several decades. As medium and long term development, core concepts have been studied for enhancing consumption of plutomum with the higher moderator to fuel ratio than the conventional fuel lattices in 100 % MOX fuel cores[l-2]. In order to measure the main core physics parameters of high moderation MOX fuel cores, an extensive experimental program, MISTRAL[3], was undertaken as a joint study of the Nuclear Power Engineering Corporation (NUPEC), CEA and its industrial partners. NUPEC has been conducting these studies on behalf of the Japanese Ministry of International Trade and Industry (MITI). This paper presents some of results of the advanced LWR core design study on plutonium high consumption MOX cores, and the status of critical experiments for high moderation MOX lattice. ' Present address: Hitach, Ltd, 7-2-1, Omika-cho, Hitachi-shi, 319-12, Japan 273 2. ADVANCED LWR CORES FOR EFFECTIVE USE OF PLUTONIUM 2.1. Design Conditions and Target of Study The study has been performed based on cores of APWR[4] and ABWR[5]. The basic specifications of those plants are shown in TABLE I and these parameters including the power densities of cores have been conserved in this study. As an index to measure the efficiency of plutonium use, "fissile plutonium consumption rate" was defined as follows: Fissile Plutonium (Puf) Consumption rate = (Loaded Fissile Plutonium - Discharged Fissile Plutonium) / (Loaded Fissile Plutonium) The target of the study is to increase this Puf consumption rate. Systematic sensitivity study of the Puf consumption rate has been done for main design features of LWR cores. The results have shown that increase of moderation ratio of the cores is most effective to increase the Puf consumption rate. 2.2 High Moderation PWR Preliminary fuel design study has been performed to increase the core moderation ratio with two methods: (1) reducing the fuel pin diameter and (2) replacing fuel pins by water rods. The influence of Table I Basic Design Parameters APWR ABWR Rated Thermal Power 4100 MW 3926 MW Effective Core Height 3.66 m 3.71m Operation Cycle Length 15.5 EFPM 15 EFPM Number of Fuel Assemblies 257 872 Fuel Assembly Type 17x17 9x9 Maximum Bum-up 55 GWd/t 55GWd/t 1.3 1.3 _E •— Reduction of Pin Diameter as °- Replacement to Water Rods Q. 1 1-2 ©1.2 cc J1.1 LI- o o 'S i £1.0 JQ I <D U) - Reduction of Pin Diameter Hi I 0.9 < • Replacement to Water Rods 1.0 0.8 3.0 4.0 5.0 6.0 7.0 4 5 6 Moderation Ratio (H/HM) Moderation Ratio (H/HM) Fig. 1 Local Power Peaking of High Fig. 2 Maximum Heat Flux of High Moderation Assemblies (PWR) Moderation Assemblies (PWR) 274 these two methods on in-assembly power distribution and thermal margin has been studied with 17 x 17 fuel assembly. Typical results, shown in Figs.l and 2, show advantage in reducing fuel diameter for increasing moderation ratio in terms of thermal margin. Thermal hydraulic margin decreases with increase of moderation ratio so that the moderation ratio should be selected properly to keep enough thermal margins. For the one of options to conserve adequate thermal hydraulic margins, a core design of the hydrogen to heavy metal atomic ratio (H/HM) of 5.0 has been studied with fuel rods of a reduced diameter. That H/HM was obtained by reducing the fuel diameter from 9.5 mm to 8.8 mm with the same fuel pin pitch. Full MOX core design with original 17x17 fuel assembly of the APWR has been also conducted for the reference of this advanced design. TABLE II shows the specifications of those fuel assemblies. The core performance is shown in TABLE III. The effect of increasing H/HM is appeared in the necessary fissile plutonium enrichment for the same operation cycle length, which decreases by 1.4 wt % corresponding 19 % reduction. On the other hand, the number of refueling assemblies increased from 88 to 108 due to the decrease of fuel inventory of the core. The maximum assembly bum-up for both cores are less than the limitation (55 GWd/t). Fig. 3 shows depletion behavior of power distribution in the core. Since full MOX core of PWRs requires no burnable absorber such as gadolinium and applies the same enrichment for all fuel N pins, peaking factors of linear heat rate (FQ) and enthalpy rise (FAH ) are well suppressed and change Table H Specificatiors ofH^jy MxieratedM)X Asserrbfy (PWR) Reference 17X17 Assembly Highly Moderated Asserrbly Moderation Ratio (BHM*) 4.0 5.0 Fuel Pin DiarrEter 9.5 mm 8.8 mm Asserrbly Lattice 17X17 17X17 (sarre pitch) Nurrber of Thimbles 25 25 Fissile HutOTumErmchrrEnt** 7.2 wt% 5.8 wt% Matrix Depleted Uranium Depleted Uranium Burnable Absorber — — *: Atomic ratio of hydrogen to heavy metal. **: Enrichment is uniform in the assembly Table HI High Moderation Core Specifications (PWR) Reference Core High Moderation Core Moderation Ratio (H/HM*) 4.0 5.0 Number of Refueling Assembly 88 108 Operation Cycle Length 15.5 EFPM*** 15.5 EFPM*** Cycle Burnup 16.5 GWd/t 19.0 GWd/t Maximum Assembly Burnup 52.7 GWd/t 53.5 GWd/t *: Atomic ratio of hydrogen to heavy metal. **: Enrichment is uniform in the assembly ***: Effective Full Power Months. 275 smoothly with depletion. The axial power distribution of full MOX cores are slightly shifted to the bottom of the cores compared with typical UO2 cores. The axial offset of full MOX cores is more negative than typical UO2 core. Such power distribution characteristics provide additional DNBR margin for the reference core and this high moderation core. Fig. 4 shows that the DNBR analysis results for the reduced diameter fuel design with practical power shapes, which add excess margin. As the fuel pin diameter is decreased for high moderation ratio, DNBR margin becomes small because of increase of heat flux and decrease of coolant flow speed caused by increase of channel flow area. These analyses show that such DNBR decrease can be 1.5 2.5 -— Reference Design Power Shape — Reference Core (H/HM=4.0) CD 1.4 <D -°-1 ncluding the Advantage for Axial —- High Moderation Core a: Power Shape (H/HM=5.0) n 1.3 io -*• I ncluding the Advantage for Axial 1 and Holizontal Power Shape •o 2.0 c a. 1.2 <u O -a o as s ff .3 1.1 O ID CO o s to Bi 1.0 r Axi -10 fo 1* as <z a> CD 0.9 Q. Q E 0.8 1c 1.0 -15 0.7 5 10 15 20 3.0 4.0 5.0 6.0 7.0 Cycle Burnup (GWd/t) Moderation Ratio (H/HM) Fig. 3 Depletion Behavior of Peaking Fig. 4 Power Distribution Effect Factors and Axial Offset (PWR) on DNBR (PWR) 60 2.5 -^-Fissile Plutonium Consumption 4.0 Rate a- Fissile Plutonium Consumption —o— Reference Core =•50 * 1 (H/HM=4.0) -»— Fissile Plutonium Loading * 2.0 3 3.5 •• V) - °- High Moderation Core c o (H/HM=5.0) .240 O T3 1.5 as (0 / §30 c 2.5 o / y 1.0 a £20 - •§2.o a 3 a. 0.5 <u 1.5 h iio u. UL 1.0 0.0 70 80 90 100 110 120 3.0 4.0 5.0 6.0 Number of RCC Moderation Ratio (H/HM) Fig. 5 Reduction of Required Fig. 6 Plutonium Balance (PWR) RCC Number (PWR) *: Capacity factor is assumed to be 100%. 276 N recovered with these power distributions. Since FAH of the full MOX cores are 2-3 % less than typical UO2, it provides about 5 % additional margin of DNBR. The axial power distribution deference also provides about 10 % additional DNBR margin. Those analyses indicate that this high moderation foil MOX core has the same DNBR margin as the reference core. For the reference full MOX core, the soluble boron worth is about 1/3 of the boron worth of typical UO2 cores.