Nuclear Fuel Optimization and Problem of Xa9953256 Increasing Burnup and Cost Efficiency of Light Water Reactors in Russia

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NUCLEAR FUEL OPTIMIZATION AND PROBLEM OF XA9953256 INCREASING BURNUP AND COST EFFICIENCY OF LIGHT WATER REACTORS IN RUSSIA

M.I. SOLONIN, Yu. K. BIBLASHCVLI, F.G. RESHETNIKOV, F.F. SOKOLOV, A.A. BOCHVAR All-Russia Research Institute of Inorganic Materials, Moscow, Russia

Abstract

Brief analysis is given of the results of WWER-1000 and WWER-440 fuel assembly (FA) and fuel rod operation to the average burn-up of 44 and 34 MW-day/kg U, respectively. The reliable operation of the fuel is noted which made it possible to outline the paths of further improvements in their fuel cycles. Based on a series of improvements reactors are being switched on a 4-year cycle; a 5-year cycle is contemplated for the WWER- 440. The fuel cycle of the boiling RBMK reactors is subjected to essential alterations. Specifically, the use of erbium oxide as an integral burnable absorber allows a higher enrichment of uranium and bum-up. To extend the burn-up a new zirconium alloy E-635 is assumed to be used as fuel rod claddings and other FA components; the alloy has higher corrosion and irradiation resistance.

INTRODUCTION

The current technical level of the PWR and BWR fuel production and the perfected system of the reactor management have allowed to provide the successful operation of PWR NPP to the average burn-up of ~45 MW* day/kg U in all countries of developed nuclear power. The differences in the reliability and FA serviceability as well as in the number of leaky fuel rods are rather small and the attention is usually not focused on this issue. More urgent and common is the tendency to the cost-efficiency of nuclear power as well as the reliability and safety of NPP. For the optimised fuel cycles of WWER NPP now in operation and new generation reactors that are under development a substantial increase of the burn-up to reach 60 MW*day/kgU is contemplated also via the use of MOX fuel. This extended burn-up makes some problems more acute, the ones that at the current burn-up were more or less noticeable. Specifically, they comprise corrosion resistance and mechanical properties of fuel rod claddings; dimensional stability of FA; pellet cladding interaction (PCI), higher FGR etc. Problems are also encountered that are primarily inherent in improving the safety as well as the cost-efficiency of boiling type reactors RBMK (BWR). The nuclear power economics is much dependent on the degree to which the nuclear fuel cycle is optimized. In Russia from the time of the nuclear power evolution the closed nuclear fuel cycle concept was adopted. To implement the concept all the needed infrastructure, the spent nuclear fuel reprocessing plant included, was set up; this made it possible to close the fuel cycle in terms of uranium 15 years ago. This in its turn substantially reduced the demand for natural uranium.

1. BRIEF ANALYSIS OF WWER OPERATION

To-day 20 units of WWER-1000 and 27 units of WWER-440 are in operation at NPP in Rus- sia and some other countries. The implemented complex of investigations aimed at improving and stabilising the quality of all fuel rod components as well as the optimised operating conditions of the reactors made it feasible to switch the WWER-1000 to a four-year fuel cycle. The mean fuel burn-up in FA discharged in 1994-1995 after operation for four cycles is some 44 MW-day/kg U. It is important to note that the number of leakers does not increase. The more so, as the experience is gained their numbers are decreasing and in 1994-1996 they were (1.2-^2.2) x 10"5. The fuel rod failures are for the most part of a "gas leakage" nature.

163 The results of the operation are adequately stable essentially for all 27 units of WWER-440. The average burn-up is 34.1 MW*day/kg U; while the maximum burn-up of 49.3 MW*day/kg U was reached by 4.4% enriched fuel of the four year cycle. The loss of tightness by fuels per cycle is on the average 0.6 xlO"5; during 1991-1995 no leakers were detected in many units. Post-irradiation examinations of FA and fuel rods irradiated to various burn-up allow one to assess the reliability and capability of the fuel rods and FA and to reveal those parameters or factors that limit a further extension of the fuel burn-up. First of all, the reliable operation of Zr-l%Nb alloy claddings is to be noted in the ammonia- boron-potassium water chemistry. The outer surface of the claddings is coated with a dark uniform oxide film 3-8 um thick; nodular corrosion is not available. At the inner cladding surface the oxide film thickness is variable and can reach 17 u,m. The claddings show a small amount of zirconium hydrides while the hydrogen content is within (3 -^6) x 10"3 % (mass). Thus, judging those factors the Zr-l%Nb alloy claddings have not yet exhausted their potential at the above burn-up. The situation is somewhat different in relation to the mechanical properties of the claddings. These properties are little dependent on the fuel burn-up up to 43-45 MW-day/kg U. The claddings retain their strength and adequate margin of ductility. During operation the outer diameter of the cladding becomes smaller because of the difference between the pressures effected by the coolant and the gas inside a fuel rod. However, with the burn-up increased to more than 45-47 MW-day/kg U the cladding diameter is increased (fig.l).

D - calculated results • - experimental results

20 30 40 Burn up, MW day / kgU FIG. 1. Fuel element diameter variation vs burn-up

This is a consequence of a gradually growing release of FGR with the fuel burn-up (fig.2) and a more intensive PCI.

calculated results experimental results

42 46 50 54 Fuel rod burn-up, MWday/kgll FIG.2. Fission gas release from WWER-440 fuel at externed burnup

164 The mean fuel rod elongation resulting from the irradiation induced growth at up to 50 MW-day/kg U burn-up obeys the relation AL=0.01xB, where AL is a mean fuel rod elongation (%), B is burn-up (MW-day/kg U). The irradiation growth of fuel rods drastically increases with the burn- up (fig.3).

10 20 30 40 Burnup, MW day/ kgU

FIG.3. Fuel element dimensional element stability

The Zr-l%Nb alloy claddings have an adequately high crack resistance. Fig. 4 illustrates the crack velocity as a function of the stress intensity factor for claddings of recrystallized Zr-l%Nb alloy and annealed Zry-4, namely, for the conditions in which those alloys are used as fuel rod claddings in commercial reactors. The comparison between the crack resistances of those two alloys in an iodine medium shows that the threshold stress intensity factor of Zr-l%Nb alloy is noticeably higher than that ofZry-4.

10 Zyrcaloy-4 (SR \ A—

1& / Zr-1%Nb (R X)

\ 10

I I 4 re .5 "e 10 1 Test temperature 350°C i 10 I I • I

•7 10 10 K,

FIG. 4.Crack propagation rate versus stress intensity factor for Zr-l%Nb cladding (%) and Zircaloy-4 (•)

165 2. PERFECTION OF WWER FUEL CYCLE AND WAYS OF BURN-UP EXTENSION

Next to the nuclear power safety the second important criterion is its cost-efficiency. It depends upon many factors. The majority of them directly related to the reactor core can be combined under the general concept, namely, perfection of the reactor fuel cycle. It is clear that each reactor is character- ized by the specific features of its fuel cycle. Currently, under improvements are fuel cycles of WWER-1000 and WWER-440.

The main efforts to optimize the WWER-1000 fuel cycle comprise the following. • Integral fuel Gd2O3 burnable absorber (IFBA) is in the process of introduction. For the purpose a special production was established where fuel assemblies were fabricated in adequately high quantities and loaded into reactors. • Steel spacer grids and guide thimbles are being replaced by Zr ones. The commercial production of those components has been mastered and currently they are in-pile tested. • Based on the above measures the reactors are being switched to the four-year cycle. • The work is under way to increase the fuel content per a fuel rod via the pellet hole reduction and optimisation of the dimensions and tolerances for the diameters of cladding tubes and fuel pellets. • The low leakage loading pattern with a longer time between reloads is being optimised.

The following is contemplated for the WWER-440 reactors: • Conversion of all reactors to a four - and then five-year cycle. • Profiled enrichment of fuel across FA.

The reactors were converted to FAs having Zr spacer grids to be operated in the four-year fuel cycle. In connection with a substantial reduction of steel in the reactor core the issue of a lower Hf content in Zr became urgent. Therefore, beginning from 1993 zirconium is commercially produced having <100 ppm Hf in place of the 500 ppm Hf content that was tolerable previously. The implementation of the above measures will result in 12-19% reduction of the specific con- sumption of fuel per unit of electricity generated and a higher cost-efficiency of NPP. The activities to introduce those improvements will be basically completed by 2000. As it has already been mentioned some of the improvements have been implemented. Besides, the WWER-440 reactors having Zr spacer grids have been switched to the four-year fuel cycle. Under consideration is the feasible conversion of those reactors to the five-year cycle. A large scope of work is under way to optimize the WWER-1000 FA. To further improve the reactor cost-efficiency and safety an alternative FA design is under developments. This FA has no steel available within the disposition of uranium oxide. The fuel assembly has a support structure fabricated from Zr-alloys including the central channel of the higher strength multicomponent E-635 alloy. This will prevent the FA distortion in a more reliable way: the fuel pellet is subject to some changes, namely, the central hole is reduced from 2.4 mm to 1.4 mm. The physical calculations indicate that 4.4% enriched fuel integrated to the Gd burnable absorber makes the four-fold reload realizable at the discharged fuel assembly averaged burn-up of 50 MW-day/kg U and the maximum burn-up of-55 MW-day/kg U. The E-635 alloy having higher corrosion and irradiation resistances is assumed to be used as fuel rod claddings and spacer grids in reactors of the 60 MW-day/kg U burn-up and the 5 year cycle.

3. RBMK FUEL CYCLE OPTIMIZATION AND HIGHER FUEL BURN-UP

Every improvement of the RBMK fuel and the core as a whole is assessed in terms of increasing its reliability and safety. These parameters are substantially dependent on the steam void coefficient of reactivity. In the reactors in operation the steam void coefficient of reactivity was lowered

166 via extra absorbers introduced into the core. However, this resulted in a substantial decrease of fuel burn-up and an increase of the FA channel power. However, the calculations demonstrate that this problem may by resolved in a more efficient way through integrating Er2O3, a burnable absorber, to fuel [2]. First, this will eliminate the extra absorber introduction. Second, absorbers available in fresh FAs reduce their power and significantly decrease a local power flash-up when FAs are loaded into a reactor. All this taken together simplifies the reload procedure and control of the power distribution within the core. A lower non-uniformity of the power density will allow a higher enrichment, thus, increasing the fuel burn-up. The production of the uranium-erbium fuel for RBMK-1000 and RBMK- 1500 has been mastered. Presently have been fabricated and currently under test are 200 assemblies with this fuel at the Leningrad NPP and 150 assemblies at the Ignalina NPP. All RBMK reactors are assumed to be converted to the uranium-erbium fuel. The erbium content of the fuel is 0.4-0.5% mass. The fuel enrichment was increased from 2.4% to 2.6% for RBMK-1000 and from 2.0% to 2.4% for RBMK-15 00. The introduction of erbium has no effect on FGR [3]. One of the distinctive features of the RBMK fuel performance is that at a substantially lower fuel bum-up compared to that of WWER, the cycle time is adequately long (1100-1200 eff. days). In this case the fuel rods operate under the conditions without control for radiolytic gases at the oxygen content of the water up to 20 u.g/dm3. However, the limitation of the used Zr-l%Nb alloy is a drastic degradation of its corrosion resistance at the oxygen content of the water above 10 p.g/dm3. This results in nodular corrosion and an aggravation of general corrosion on the waterside, particularly, under spacer grids. This in its turn leads to claddings being superheated, local hydrogen uptake etc. The sec- tion of fuel rod cladding under a spacer grid may be decreased by 40%. Those limitations may be obvi- ated via the application of the well-studied Zr-l%Nb-1.3%Sn-0.35%Fe alloy as a cladding material. This alloy is much superior to the Zr-l%Nb one not only in irradiation but also in corrosion resistance and which is very important it is not prone to nodular corrosion. This is particularly attractive for the claddings of boiling reactor fuel rods, operating without control for radiolytic gases. Fuel rods clad in this alloy were representatively tested at the Leningrad NPP under the standard operating conditions. The post-irradiation examinations have shown that the fuel rod claddings are coated with a lustrous uniform dark oxide film. No traces of nodular corrosion are available. The local corrosion of the fuel rod claddings under the spacer grids is some 3.5 times lower than that of Zr-l%Nb alloy. Russia has adopted the concept of the closed nuclear fuel cycle. To implement this concept among other things a plant of the reprocessing capacity 4001 spent fuel/per year was set up [4]. This made it feasible to close the fuel cycle for U. Of the variety of the fuel rods used in the nuclear power of Russia only the reprocessing of RBMK spent fuel rods is not contemplated. They are shipped to be long-term stored. Their low residual content of U-235 and the low content of fissile spe- cies in the generated plutonium make their reprocessing cost inefficient. Meanwhile, the RBMK reactors are actively involved in the closed nuclear fuel cycle since their fuel is for the most part fabricated from recovered uranium after adjustments for the U-235 content are made. The use of the recovered uranium has not essentially affected the fuel rod production process or the reactor performance. The recovered uranium is also assumed to be used in the WWER reactors. This significantly reduces the demand for natural uranium.

References

[1]. Solonin, M., Bibilashvili, Yu., Medvedev, A., Sokolov, F., et al. "LWR fuel performance and material development activities for extended burnup in Russia", paper presented at the 1997 Interna- tional Topical Meeting on LWR Fuel Performance, Portland, Or., March (1997). [2]. Panushkin, F., et al. Adding erbium to increase RBMK safety, Nuclear Engineering Interna- tional. Nov. 1995. [3]. Bibilashvili, Yu., Reshetnikov, F., Ioltukhovsky, F., Sokolov, F., et al., "Status and development of RBVK fuel rods reactor materials", Water Channel Reactor Fuel (proc. IAEA Tech. Comm. Mtg. Vienna, 1996), IAEA-TECDOC-997, Vienna (1998) 151-165. [4]. Bibilashvili, Yu., Reshetnikov, F., et al. "Russia's nuclear fuel cycle: an industrial perspec- tive", IAEA Bulletin, vol.35, N3, 1993. Vienna, Austria.

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