Comprehensive Study of Lattice Cell Calculations for Thorium Based Fuel Cycle in Light Water Reactors Using SRAC Code

Transcription

1 Comprehensive Study of Lattie Cell Calulations for Thorium Based Fuel Cyle in Light Water Reators Using SRAC Code Le Dai Dien, Ha Van Thong and Giang Thanh Hieu Institute for Nulear Siene and Tehnology, INST 5T-160 Hoang Quo Viet, Nghia Do, Ha Noi-VietNam The designers of the innovative reators have proposed a number of approahes to inreasing resoure effiieny. Adding thorium, a fertile material, to the fuel is onsidered in this report. Under this approah, a large portion of the reator output is produed by fissioning of the 233 U resulting from neutron apture by thorium, whih results in redued requirements for naturally-ourring fissile uranium ( 235 U). The proliferation potential of the light water reator fuel yle may be signifiantly redued by utilization of thorium as a fertile omponent of the nulear fuel. The onept of using Th- 233 U as fuel has been applied to an existing LWR design as ompare with another fuel yles (UO2 and MOX). SRAC ode is extensive used to investigate the lattie ell problem. 1. Introdution. Nulear fuels used in reators an be 235 U, 239 Pu and/or 233 U. The ontent of natural uranium ontains 99.3% 238 U, 0.7% 235 U. The fuel irradiated in onventional light water reators must be enrihed from 2 to 4% 235 U to maintain nulear fission hain reation by using light water as the moderator and oolant. 239 Pu an be produed from fertile nulide 238 U, while 233 U produed from 232 Th. Thorium is muh more abundant in the earth s rust than uranium and the need of Pu burning from existing Pu stokpiles make the thorium based fuel yle is widely onsidered for many deades. Thorium like uranium an be used as fuel in nulear reators though thorium is not fissile material, but 232 Th is apable to apture slow neutrons to form 233 U, a fissionable isotope. Thus, thorium based fuel yle an be used in all proven reator types (1). 233 U produed from 232 Th in the neutronis point of view is one of the best isotope in the fissionable isotopes. In all energy range, neutron fission yield ratio (η) and the number of neutron absorbed are higher than those of 235 U and 239 Pu, so that 233 U ould be used as fuel for many kind of reator. Thorium oxide ThO 2 has greater stability and an be used with high temperature, longer durability due to its melting point of C (UO 2 : C) that expeted to gain high burn-up. The reator fuelled by thorium will not reah ritial but it an use a mixed ore as the seed-and-blanket onept. 233 U would be produed whih in turn fuel either the initial reator. This feature is expeted to be used to onsume a large plutonium stokpiles today. One of the advantages of 233 U as ompare with 235 U and 239 Pu is that the higher neutron emitted yield when one neutron absorbed. The 235 U or 239 Pu are used to breed fissionable isotope 1

2 from thorium. 232 Th absorbs neutron to beome 233 Th and then deay to 233 Pa and finally to 233 U by deay hain: The fuel is irradiated in the reator ore, in the bak end of fuel yle 233 U an be extrated from thorium and reused as fuel to make a lose yle. In this study, from reator physis alulation aspet, our work fouses on estimation of nulear fuel onversion of 233 U and those of uranium or MOX fuel yles. Capture ross setion of 232 Th and apture and fission ross setions of 233 U give the fuel onversion ratio. The ratio of the number of fissionable nulei produed from fertile material to the number of fissionable nulei onsumed in fission and non-fission reations. It is given by: Fer Fer Fis Fis Fis N σ = N ( σ + σ ) Thus, ratio of fissionable ontent and fertile ontent in the fuel an be defined by: Fis Fer N σ = Fer Fis Fis N σ + σ Fer σ In thorium based fuel yle, ratio of ross setions Fis σ f + σ reproduing of 233 U fuel during the fuel irradiated in the reator ore. f f Fis identifies the apability of For a purpose of proving the advantages of the thorium fuel yle, the preliminary reativity alulations were performed for latties of fuel rods ontaining ThO 2 and (Th,U)O 2 as well as UO 2 and MOX. The reator would be water ooled and retains all design features of a LWR. 2. Lattie ell alulations with LWRs. The lattie ell of LWRs has a pin ell formed a square lattie as in Figure 1. The geometry parameters are in Table 1. Parameters R1(mm) R2(mm) L(mm) Table 1: Lattie ell parameters of LWRs. Material (temperature) Fuel (900 K) Clad (600 K) Water (600 K) PWR BWR From physis alulation aspet for lattie ell problem, the differene between PWR and BWR is that the oolant in PWR is not boil while in BWR the vapor ontent an be up to 40% in 2

3 normal operation. The ontent of UO 2 3% -ThO 2 97% is used for Thorium fuel pin and UO 2 4% used for PWR and BWR pin ell alulations. Figure 1: Lattie ell onfiguration of LWRs. Figure 2: K-inf vs. Fuel Burn-up. The variation of multipliation fator K-inf on fuel burn-up is presented in Figure 2. MOX and Th/ 233 U fuel an be burn up to 60GWd/T, while UO 2 fuel an be reah maximum at 35GWd/T the average burn-up of present LWRs. 3. Estimate of fuel onversion fator. With UO 2 and MOX fuels the onversion ratio is in the range of 0.5 to 0.7. Espeially in MOX fuel, due to the main fission isotope is 239 Pu so that the onversion fator is not so high, while the 232 Th/ 233 U fuel it is muh higher as 233 U is breeded during fuel irradiation in the ore, the value obviously is greater than 1. The Figure 3 illustrates the variane of onversion fator by fuel burn-up of three fuel yles. It should be noted that if the UO 2 /MOX fuels are used in ombination with Th/ 233 U in a ertain onfiguration of the ore, the fuel onversion ratio would be improved, it will make the fuel burn-up more higher and save the 235 U fuel as well as the good option for onsumption of plutonium. 3

4 Figure 3: Conversion ratio (C.R) vs. Burn-up. Further, from Table 2 we an see that the 239 Pu ontent produed from Th/ 233 U yle is muh less than those formed in UO 2 or MOX fuel yles. This is also an advantage of thorium based fuel yle as plutonium an not be extrated during reproessing. GWd/T Table 2: 239 Pu ontent in the fuel burn-up. PWR BWR UO 2 Th- 233 U MOX UO 2 Th- 233 U MOX 1.0E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E-04 In general, Th/ 233 U fuel is used in ombination with UO 2 fuel or MOX in fuel assemblies. The pratial use has been investigating in India and many modeling studies have also been applied to present exist designs as PWR, VVER, PHWR (4-9). 4

5 In this study, two kinds of fuel are ombined into one fuel rod. The UO 2 fuel as seed is entered and surrounded by Th/ 233 U fuel outside. Figure 4: Two-layer fuel pin onfiguration. The Figure 5 represents the dependene of multipliation fator K-inf and onversion ratio at the burn-up of 40 GWd/T by the volume ratio of UO 2 fuel alloy (V F ) and Th/ 233 U fuel (V T ). When the UO 2 volume inreases, multipliation will be inreased and onversion ratio dereases. However, these parameters will have minor hange in the range 1.5 and 2.5 of the ratio V F /V T values. It should be noted that two-layered fuel rod may be one option in hoosing the (4, 6, 7) onfigurations of fuel assemblies beside well known onfigurations that have been investigated. The detail investigations should be arried out to onfirm the feasibility and appliability of this onfiguration. Figure 5: The multipliation fator K-inf and onversion ratio at burn-up of 40 GWd/T vs. Volume ratio of two-layer fuel rod. 4. Conlusion. In the next several deades, the onventional LWRs based on uranium fuel yle are still used. However, the other types of reator are under extensive development, the typial reators are 5

6 FBR, FBMR, HTGR et. With the emphasize on transuranium and MA transmutation, the Th/ 233 U fuel yle with the important advantages: + Contribute into burning of plutonium stokpiles, and 239 Pu produed by this fuel yle is muh less than MOX or UO 2 fuel. + High radioative waste with large lifetime is less than other fuel yles. + High fuel burn-up. + Used for high temperature reators (HTR). + And sustainable as ompare with limited uranium resoure. It will definitely be one of the remarkable options for nulear fuel yle in the future. Referenes. [1] Thorium based fuel options for the generation of eletriity: Developments in the 1990s. (Vienna, IAEA-TECDOC-1155, May 2000). [2] K. Okumura, T. Kugo, K. Kaneko, K. Tsuhihashi. SRAC: The omprehensive neutronis alulation ode system. JAERI/Code [3] Le Dai Dien, Ha Van Thong, Giang Thanh Hieu. Appliation of Pij module in SRAC ode system for lattie ell problems of PWR and BWR Report, CS/03/ VAEC. [4] K. Okumura, H. Unesaki, T. Kitada, E. Saji - Benhmark results of burn-up alulation for LWR next generation fuels. PHYSOR 2002, Seoul, Korea, Otober 7-10,2002. [5] D.F.Torgerson, P.G. Bozar, A.R.Dastur. CANDU Fuel Cyle Flexibility. AECL (1994). [6] P.S.W.Chan, P.G.Bozar, R.J.Ellis, F.Ardeshiri. Fuel management simulations for one-through thorium fuel yle in CANDU reators. IAEA Tehnial Commettee Meeting on Fuel Cyle Options for LWRs and HWRs, [7] K. Balakrishinan. HWR advaned fuel yle flexibility and sustainable development, Ibid. [8]. D.E. Beller, W.C. Sailor, F.Venneri. A losed ThUOX fuel yle for LWRs with ADTT(ATW) bakend for 21 st entury. LA-UR (1998). [9] A.Yamamoto, T.Ikehara, T.Ito, E.Saji. Benhmark Problem suite for reator physis study of LWR next generation fuels. J. Nul. Si. Tehnol., 39(8), 900 (2002) 6