Conservative approach for PWR MOX Burnup Credit implementation

Transcription

1 International Conference on the Physics of Reactors Nuclear Power: A Sustainable Resource Casino-Kursaal Conference Center, Interlaken, Switzerland, September 14-19, 2008 Conservative approach for PWR MX Burnup Credit implementation Ludyvine Jutier a, *, Benoît Checiak a, Jérôme Raby a, Luis Aguiar a, Igor Le Bars a a IRSN, Fontenay-aux-Roses, France Abstract Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumption made to: define the axial profile of the burnup, determine the composition of the irradiated fuel and compute the criticality simulation. In the framework of Burnup Credit implementation for PWR mixed oxide fuels (MX), this paper focus on the determination of a conservative inventory of the irradiated fuel. The studies presented in this paper concern: the influence of irradiation conditions and of the MX fuel initial composition on the irradiated MX fuel reactivity. Criticality calculations are also performed for PWR MX fuel industrial applications in order to get Burnup Credit gain estimations. 1. PWR MX Burnup Credit context Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for nuclear fuel cycle applications. In the early 80 s, a method was accepted by the French Safety Authorities to take Burnup Credit into account for reprocessing, storage and transport with very pessimistic hypothesis: only actinides were considered and the amount of burnup used in the studies was equal to the mean burnup in the 50- least-irradiated centimetres of the fuel assembly (Maubert, 1987). As many facilities want to optimize their processes, IRSN (IPSN at this period), CEA and the main French nuclear companies set up a working group in order to study the different ways Burnup Credit could be taken into account considering some fission products and a burnup axial profile. Today, the main studies of the French Burnup Credit Working Group on PWR UX fuels have been performed (Raby, 2003). In these studies, conservatisms have been reduced with respect to the 50-least-irradiated centimetres and actinide-only method; then it is necessary to carefully analyze the validity of the assumptions made to: define the axial profile of the burnup, determine the composition of the irradiated fuel and compute the criticality simulation. A Burnup Credit implementation for PWR mixed oxide fuels, i.e. fuel containing a mixture of uranium and plutonium oxides (MX), requires the same analysis. This paper proposes to focus on the determination of a conservative inventory of the irradiated fuel. ne of the main difficulties of a burnup credit implementation for PWR MX fuel is the wide range of parameters compared to PWR UX fuel * Corresponding author, ludyvine.jutier@irsn.fr Tel: ; Fax:

2 (plutonium composition, plutonium content, uranium composition, presence of 241 Am due to the 241 Pu decay, zoning of assemblies). Actually, it is easy to determine a conservative composition of the fresh fuel for PWR UX Burnup Credit applications to perform criticality studies: for a given burnup, the maximal initial uranium enrichment leads to the most reactive spent fuel for all applications. But for PWR MX fuels, a fresh fuel with a conservative isotopic vector (of plutonium) will not necessarily lead to the most reactive fuel after irradiation: the fissile isotopes of plutonium will disappear and give absorbent isotopes, while captures on absorbent isotopes of plutonium will give new fissile isotopes Therefore, to simplify MX burnup implementation, one way is to set up a method to determine, for a given ratio of Pu/(U+Pu), a bounding plutonium vector for the fresh fuel that gives, after irradiation, the most reactive fuel, whatever the irradiation history (Raby, 2007). Indeed, the study pointed out that, for long cooling times, the bounding initial plutonium vector would minimize the amount of 241 Pu, and for short cooling times, the bounding vector would minimize or maximize the amount of 241 Pu depending on both the burnup value and the weight ratio Pu/(U+Pu). The studies presented in this paper concern: the influence of irradiation conditions on MX fuel reactivity, criticality calculations for MX assemblies storage configurations and transport package, and the evaluation of the conservatism of the fresh fuel bounding plutonium vector method. The bounding plutonium vector analyses are also completed by studying the effect of 238 Pu and 242 Pu presence in the fresh MX fuel (these isotopes were not initially taken into account in the method). 2. Calculation method and models 2.1. Depletion calculations The fuel inventory after irradiation of the MX assembly in a pressurised water reactor is computed by the French DARWIN 2.0 system (Tsilanizara, 2003) based on APLL2 and PEPIN 2 depletion codes. The APLL2 assembly calculation solves the integral form of the Boltzmann equation through the collision probability method. The PEPIN 2 evolution module then uses the results provided by APLL2 to make up the collapsed library with burnup dependent cross-sections required to characterize the isotopes described in the depletion chains. For this depletion calculation, an infinite array of PWR AFA type MX assemblies (3 zones of plutonium content) for different MX fuel compositions and with or without control rods inserted is considered Criticality calculations The fuel inventories, coming from the depletion calculations, are used in the calculations of the neutron multiplication factor (k eff ) of an assembly surrounded by 20 cm of water, different configurations of storage and transportation. These criticality calculations are performed with the standard route of the criticality-safety package CRISTAL V1 (Gomit, 2003). This calculation uses the APLL2 computer code to generate selfshielded cross sections from 172 energy-groups libraries, and the multigroup Monte Carlo computer code MRET 4 for k eff calculations. For those calculations, different axial profiles of burnup are used: a flat profile, a standard PWR profile and a penalizing profile disturbed by a control rod insertion during irradiation (Maillot, 1999). These profiles established for PWR UX are considered in a first approximation representative of PWR MX. As for the irradiation conditions, different histories are studied. The nuclides considered are the following 16 actinides: 234 U, 235 U, 236 U, 238 U, 238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu, 237 Np, 241 Am, 242m Am, 243 Am, 243 Cm, 244 Cm and 245 Cm, and the following 15 fission products: 95 Mo, 99 Tc, 101 Ru, 103 Rh, 109 Ag, 133 Cs, 143 Nd, 145 Nd, 147 Sm, 149 Sm, 150 Sm, 151 Sm 152 Sm, 153 Eu, and 155 Gd (Roque, 2002, Connor and Hong Liem, 2003). The steps of the calculation tools are described in Fig. 1. 2

3 Conditions Irradiation d irradiation conditions Historique Irradiation d irradiation history Description Geometry de la géométrie description A P L L 2 P E P I N 2 A P L L 2 M R E T 4 k eff (CEA 93) (CEA 93) TC Σ C(X) Σ Depletion calculations with DARWIN Calculs d évolution avec le formulaire DARWIN Criticality calculations with CRISTAL Calculs de criticité avec le formulaire CRISTAL Fig. 1. Steps of the calculation tools 3. Influence of irradiation conditions In order to study the influence of irradiation conditions for PWR MX burnup implementation, a parametric study of the effect of the main irradiation parameters is realised for a given MX fuel composition a. These parameters value is varied around numbers representative of the standard fuel management of French reactors, named nominal irradiation conditions afterwards. The range of the studied values for the different parameters are given in Table 1. The influence of each parameter is studied separately as a function of the burnup (from 0 to 60 GWd/t) and the cooling time (from 0 to 50 years) Effect of control rods French PWR operating conditions can involve periods of partial control rods insertion. In order to bound the reactivity effect due to this insertion, a full control rod insertion is considered during the entire irradiation. a Pu/(U+Pu) = 4.5 %, 238 Pu = 0.80 %, 239 Pu = %, 240 Pu = %, 241 Pu = 7.60 %, 242 Pu = 2.90 %, 241 Am = 1.30 %, 235 U = 0.72 %. Table 1 Studied irradiation conditions Parameter Nominal value Range Moderator temperature Moderator density Fuel temperature Boron concentration Specific power 310 C C C C 550 ppm ppm 41 W/g W/g Control rods ut ut - In The presence of control rods during the irradiation of a MX assembly induces the capture of thermal neutrons and thus the hardening of the neutron spectrum. The spectrum hardening due to the presence of control rods leads to increase the fissile plutonium content and consequently to significantly increase the reactivity: 6500 pcm for an infinite array of MX assemblies irradiated at 60 GWd/t, without cooling time. 3

4 3.2. Effect of the moderator temperature The increase of the moderator temperature reduces the moderating power of water (due to the decrease of the density) that induces, like in the previous case, the hardening of the neutron spectrum and leads to an increase in reactivity of the burned fuel. Fig. 2 shows the impact of the moderator temperature on the reactivity of an infinite array of MX assemblies for different burnup values. For assemblies irradiated at 60 GWd/t, this impact reaches 90 pcm per C. k inf (moderator temperature) - k inf (310 C) (pcm) GWj/t 50 GWj/t 40 GWj/t 30 GWj/t 20 GWj/t 10 GWj/t Moderator temperature ( C) Fig. 2. Effect of the moderator temperature on the reactivity of an infinite array of MX assemblies 3.3. Effect of the boron concentration In the same way as the presence of control rods, the increase of the boron concentration induces the capture of thermal neutrons and consequently the hardening of the neutron spectrum. The increase in reactivity of the burned fuel, due to the increase of the boron concentration, worths 1.69 pcm per ppm of bore Effect of the fuel temperature The increase of the fuel temperature induces a broadening of the 238 U capture resonance that leads to increase the neutron capture and thus the forming of fissile material following the chain reaction. The increase in reactivity of the burned fuel, due to the increase of the fuel temperature, worths 3.38 pcm per C Effect of the specific power Increasing the specific power is like decreasing the irradiation time. Since the decrease of the irradiation time reduces the forming by decay of some absorbent actinides or fission products, the reactivity of the burned fuel increases. This increase worths pcm/(w/g) Summary Conditions that maximize the reactivity of the burned fuel are: the presence of control rods during the irradiation, a maximized moderator temperature, boron concentration, fuel temperature and specific power. These conditions will be called penalizing irradiation conditions afterwards. The penalty due to control rods insertion accounts for 65 % of the total penalty. The one due to the moderator temperature accounts for 20 % and the one due to the boron concentration for 10 %. This distribution of penalties looks like the one obtained in the case of PWR UX fuel. 4. Burnup Credit gain estimation The Burnup Credit gain is defined as the difference of reactivity between an irradiated fuel and the same fuel fresh. Calculations for different configurations of storage (storage pool configuration with a borated basket containing 9 fuel assemblies) and transportation (transport cask with 13 MX assemblies) are realised for a given MX fuel composition (same as in section 3) to get Burnup Credit gain estimations for PWR MX fuel industrial applications Influence of irradiation conditions In order to study the influence of irradiation conditions, calculations are performed for both nominal and penalizing irradiation conditions defined in section 3. A flat burnup profile is considered for these calculations. The obtained values of Burnup Credit gain are presented in Fig. 3 as a function of burnup (no cooling time). 4

5 Burnup Credit gain (pcm) Nominal irradiation conditions Transport cask Fallen assembly ff-centered assemblies Higher standing assemblies Penalizing irradiation conditions Transport cask Fallen assembly ff-centered assemblies Higher standing assemblies Burnup Credit gain (pcm) Configurations Fallen assembly ff-centered assemblies Higher standing assemblies Cooling times 50 years 5 years 0 day Burnup (GWd/t) Burnup (GWd/t) Fig. 3. Burnup Credit gain for PWR MX fuel concrete applications for nominal and penalizing irradiation conditions For an irradiation at 60 GWd/t, the penalty is maximum but it represents only one third of the nominal Burnup Credit gain which gives a mean Burnup Credit gain of pcm for the different configurations under the penalizing irradiation conditions. A study of the origin of this Burnup Credit gain shows that the gain due to fission products is the same under both nominal and penalizing irradiation conditions, whereas the gain due to actinides is less important under penalizing irradiation conditions than under nominal irradiation conditions. Indeed, penalizing irradiation conditions induce the hardening of the neutron spectrum and thus the forming of fissile actinides which decreases the Burnup Credit gain due to actinides. So under the penalizing irradiation conditions, the Burnup Credit gain mainly comes from fission products. As a comparison, in the case of PWR UX fuel, the Burnup Credit gain mainly comes from actinides (Raby, 2003) Influence of cooling time Calculations performed for different cooling times indicate that the penalty due to the penalizing irradiation conditions doesn t vary very much with cooling time. Moreover, the Burnup Credit gain of the studied configurations increases with cooling time (Fig. 4). For an irradiation at 60 GWd/t, the difference of Burnup Credit gain is equal to 5600 pcm, between 0 and 5 years and equal to pcm between 0 and 50 years. Fig. 4. Burnup Credit gain for PWR MX fuel concrete applications for different cooling times 4.3. Influence of the axial profile of burnup Until now, the value of the burnup which was applied to the whole length of the fuel assembly was equal to the mean value over the fuel assembly. Because this approach is not always conservative, two different axial profiles of burnup are studied: a standard PWR profile and a penalizing profile disturbed by a control rod insertion during irradiation (Maillot, 1999). The difference of reactivity between a flat profile and an axial profile is called end-effect in this paper. Results show that the configuration for which the consideration of a burnup profile has the larger impact is the storage configuration where assemblies are standing higher than the borated basket. Indeed, in this case, the extremities are no longer separated by boron steel plates which generates interactions between assemblies. Moreover, the consideration of a penalizing burnup profile is more conservative than the consideration of a standard burnup profile. In fact, since the top of the assembly is less burned, it is more reactive than the standard profile. Burnup Credit gain and end-effect obtained with the various profiles are given in Table 2. In the worst case (higher standing assemblies and penalizing profile), the end-effect represents 43 % of the Burnup Credit gain with a flat profile. 5

6 Table 2 Influence of the axial profile of burnup Configuration Burnup Credit gain b Flat profile Standard profile End-effect c Penalizing profile Fallen assembly d 2925 ff-centered assemblies Higher standing assemblies Transport cask b k g = k eff (burned, flat profile)-k eff (fresh). c d k ee = k eff (burned, profile)-k eff (burned, flat profile). Non significant value 5. Influence of the MX fuel initial composition For PWR MX fuels, a fresh fuel with a conservative isotopic vector of plutonium will not inevitably lead to the most reactive fuel after irradiation. In practice, this implies to perform a depletion calculation for each initial MX fuel composition to use MX burnup in criticality studies. Due to the wide variety of MX fuel assemblies, this solution seems unachievable without huge calculation efforts. Therefore, a way to determine, for a given ratio of Pu/(U+Pu), a bounding plutonium vector for fresh MX fuel was investigated (Raby, 2007). nly 3 isotopes ( 239 Pu, 240 Pu and 241 Pu) were initially taken into account for the fresh fuel bounding plutonium vector. The work done shows that for commonly used plutonium (plutonium coming from the reprocessing of commercial UX fuel assemblies): - the configuration leading to the maximum neutron multiplication is always obtained with a minimum amount of 240 Pu; - the determination of bounding initial isotopic vectors of the plutonium mainly depends on the spent fuel cooling time, mainly due to the beta-decay of the 241 Pu (whose half-life is 14.4 years) in 241 Am, a neutronic absorber for thermal spectrum. Complements provided in this paper concern the impact on the above conclusions of the 238 Pu and 242 Pu presence in the fresh MX fuel and of the irradiation history Impact of the 238 Pu and 242 Pu presence in the fresh MX fuel The impact of the 238 Pu and 242 Pu presence in the fresh MX fuel needs to be checked because the 238 Pu α decay leads to the production of 234 U (stable isotope with a capture section less important than the 238 Pu one) and the 242 Pu decay leads to a higher content of curium for high burnup, which can induce an increase in reactivity. This impact is checked with calculations realised on a sample of MX fuel compositions representative of MX used in French reactors (see Table 3) for the nominal irradiation conditions. Depletion calculations are performed for each of these three MX fuel compositions and for the two bounding MX fuel compositions resulting from the application of the method described in (Raby, 2007) (see Table 3). The obtained results for different burnups and a cooling time equal to 0 are presented in Fig. 5. Table 3 Studied MX fuel compositions: MX i are real compositions representative of MX used in French reactors, F j are the associated fictive bounding compositions resulting from the application of the method described in (Raby, 2007) Pu/(U+Pu) 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu 241 Am 235 U MX % 0.30 % % % 4.20 % 1.20 % 1.20 % 0.25 % MX % 1.60 % % % 9.00 % 5.50 % 1.30 % 0.25 % MX % 1.00 % % % 7.10 % 3.30 % 1.00 % 0.25 % F % 0.00 % % % 3.74 % 0.00 % 0.00 % 0.25 % F % 0.00 % % % 8.45 % 0.00 % 0.00 % 0.25 % 6

7 International Conference on the Physics of Reactors Nuclear Power: A Sustainable Resource Casino-Kursaal Conference Center, Interlaken, Switzerland, September 14-19, 2008 k eff 0, , ,74000 F1 F2 MX 1 MX 2 MX 3 Table 4 Burnup Credit gain estimations for fuel MX 1 for the penalizing irradiation conditions Cooling time Burnup 40 GWd/t 60 GWd/t 0, , Burnup (MWd/t) Fig. 5. Effect of the 238 Pu and 242 Pu presence in the fresh MX fuel Whatever the burnup is, the reactivity of the three MX fuel compositions studied is bounded by the reactivity of the two MX fuel compositions resulting from the application of the method. The same observation can be made for the different studied cooling times. Moreover, the penalty associated with the determination of a bounding initial plutonium vector for fresh MX fuel decreases with burnup and cooling time Impact of the irradiation history The independence of the method regarding the irradiation history is checked by performing the calculations described in section 5.1 but for the penalizing irradiation conditions. Results show that the reactivity of MX 1, MX 2 and MX 3 is still bounded by the reactivity of F1 and F2 which allows verifying the conservatism of the method described in (Raby, 2007). In addition, the penalty associated with the method is less important for the penalizing irradiation conditions than for the nominal irradiation conditions. Thus, penalties due to irradiation conditions and to bounding plutonium vector for fresh MX fuel don t add up. Burnup Credit gain estimations for fuel MX 1 are given in Table 4 for the penalizing irradiation conditions. e f 0 day 50 years k r e k m f k r k m k r = k eff (burned MX 1)-k eff (fresh MX 1). k m = Max{k eff (burned F1),k eff (burned F2)}-k eff (fresh MX 1). For an irradiation at 60 GWd/t and for a cooling time equal to zero, Burnup Credit gain after the application of the bounding plutonium vector for fresh MX fuel method is around 8000 pcm which is still a significant gain. 6. Conclusion The study presented in this paper contributes to the determination of a conservative inventory of the irradiated fuel for PWR MX Burnup Credit implementation. The analysis of the influence of irradiation conditions show that conditions that maximize the reactivity of the burned fuel are: the presence of control rods during the irradiation, a maximized moderator temperature, boron concentration, fuel temperature and specific power. Criticality calculations for PWR MX fuel industrial applications allow getting Burnup Credit gain worth estimations (mainly due to fission products). Finally, calculations performed for different irradiation conditions on a sample of MX fuel compositions representative of MX used in French reactors allow to verify the conservatism of the bounding initial plutonium vector determination method (Raby, 2007) and to evaluate the associated penalty. This penalty decreases with burnup and cooling time. Further studies on PWR MX Burnup Credit implementation should focus on the definition of the axial profile of the burnup and to the computing of the criticality calculation. 7

8 References Gomit, J.M. et al., CRISTAL V1: Criticality Package for Burnup Credit calculations, Proc. of International Conference on Nuclear Criticality Safety, ICNC 2003, Tokai Mura, Japan. Maillot, M. et al., Search for an Envelope Axial Burn-up Profile for Use in the PWR Criticiality Studies with Burn-up Credit, Proc. of International Conference on Nuclear Criticality Safety, ICNC 1999, Versailles, France. Maubert, L., The Burn-Up consideration in the criticality safety of irradiated LWR fuel cycle plants, Proc. of International Seminar on Nuclear Criticality Safety, ISCS87, Tokyo, Japan. Connor, G.J., Hong Liem, P., Burnup Credit Criticality Benchmark Phase IV-B: Results and Analysis of MX Fuel Depletion Calculations, Nuclear Science NEA/NSC/DC(2003)4 ISBN Raby, J. et al., Current studies related to the use of Burnup Credit in France, Proc. of International Conference on Nuclear Criticality Safety, ICNC 2003, Tokai Mura, Japan. Raby, J. et al., Bounding plutonium vector for PWR-MX Burnup Credit applications, Proc. of International Conference on Nuclear Criticality Safety, ICNC 2007, Saint Petersburg, Russia. Roque, B. et al., Burnup credit calculation route for PWR MX assemblies and experimental validation in Minerve R1 MX and SLB1 P.I.E., Proc. of the IAEA Technical Committee Meeting on the requirements, practices and developments in burnup credit applications, Madrid, Spain. Tsilanizara, A. et al., DARWIN: an evolution code system for a large range of applications, J. Nucl. Sci. Technol., Supplement 1, pp