Plutonium multi recycling in pressurised water reactors of the EPR type using laser isotope separation of Pu242

Annals of Nuclear Energy 129 (2019) 399–411

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Plutonium multi recycling in pressurised water reactors of the EPR type using laser isotope separation of Pu242 Noel Camarcat a,⇑, David Lecarpentier b, Franck Lavaud b, Pascal Lemaire c a

EDF DIPNN, 22-30 Avenue de Wagram, 75382 Paris Cedex 08, France EDF R&D, 7 Boulevard Gaspard Monge, 91120 Palaiseau, France c CEA/MARCOULE, BP 17171, 30207 Bagnols-sur-Ceze Cedex, France b

a r t i c l e

i n f o

Article history: Received 28 October 2018 Received in revised form 1 February 2019 Accepted 4 February 2019

Keywords: Plutonium MOX fuel EPR reactors Plutonium multirecycling Plutonium single recycling in PWR’s

a b s t r a c t Plutonium multi recycling in a thermal spectrum with laser separation of even isotopes is not a new idea. It has been proposed by previous authors (Forsberg, 2015). They point out that even isotopes build-up when recycling in a thermal spectrum may create reactivity control challenges and parasitic neutron absorption. They also examine the effects induced by separation of isotope Pu240 on higher actinides production. A recent CEA internal report (Bréchet, 2017) mentions Pu242 laser isotope separation as a scientific possibility for plutonium recycling in PWR’s, with a standard thermal neutron spectrum. We present in this paper a theoretical plutonium AVLIS process. The acronym AVLIS stands for Atomic Vapor Laser Isotope Separation. We apply such a process to Pu242 depletion when MOX fuel comes out of the reactors to overcome even Pu isotopes build up and eliminate reactivity control problems in a thermal spectrum. We calculate quantitative results for a long term nuclear fleet consisting of PWR’s of the EPR type with UOX and MOX fuel and of the following fuel cycle facilities: classical reprocessing and MOX fabrication plants and theoretical laser isotopic depletion plants. Results show that the backlog of used MOX fuel accumulated during the operation of the thermal fleet between 1987 and 2050 (of the order of 5000 tons) can be processed in about 25 years between 2050 and 2075 and recycled in the PWR fleet in the 2050–2100 time frame. The isotopic tails with a mixture of the 5 plutonium isotopes can be reconditioned in a high content ceramic matrix and stored in a geological repository, without creating an increase in repository space too difficult to manage compared to the vitrified wastes already produced at the fuel reprocessing plant. Alternatively, the tails could undergo intermediate storage pending the arrival of fast neutron reactors to transmute them. Ó 2019 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY license (http:// creativecommons.org/licenses/by/4.0/).

1. Introduction and historical background Atomic Vapor Laser Isotope Separation (AVLIS) applied to plutonium is mentioned in the United States as early as 1984 in the open literature. An environmental impact statement for a complete plutonium plant located in the state of Idaho is filed by DOE in 1988 (Final Environmental Impact Statement, 1988). Price tags are even given in the regional press. Various applications of an AVLIS process to plutonium are mentioned in a US academy of sciences report in 1991, in particular Pu238 isolation for thermo-electricity generation and space applications (Alternative Applications of Atomic Vapor Laser Isotope Separation Technology, 1991). But multi-recycling of plutonium in light water reactors is not listed.

⇑ Corresponding author.

Probably since it was not a problem for the American nuclear fleet which was not using mixed oxide of plutonium and uranium fuels (MOX) in those years. Following the European approach, the French fleet has started loading MOX fuel assemblies in Saint Laurent des Eaux in 1987. The first EDF internal report about plutonium multirecycling in N4 reactors is written in 1989 (Bloch and Tetart, 1989). They clearly identify the limit of even isotopes content in plutonium. A later analysis explains the following reactor physics: Loss of water in a thermal reactor induces a hardening of the neutron spectrum. When the percentage of void in the coolant exceeds 80%, the neutron spectrum is flattened in the region 0.001 eV–100 eV and shifts to the right of the giant resonance peaks of Pu242 (centered at 2,7 eV) and of Pu240 (centered at 1,0 eV). See the normal and voided flux in Fig. 1 (the voided flux has only a fast component around 1 MeV). Neutron absorption no longer occurs in those resonance

E-mail address: [email protected] (N. Camarcat). https://doi.org/10.1016/j.anucene.2019.02.010 0306-4549/Ó 2019 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY license (http://creativecommons.org/licenses/by/4.0/).

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Fig. 1. Neutron spectrum, nominal and fully voided situations. Apollo2 neutronics code calculations.

peaks leading to an important reactivity increase. This is called the Water Void Reactivity Effect (effet de vidange in French), designated by the symbol dq. Positive values of dq have to be excluded for safety considerations. Practical engineering limits to avoid positive dq’s for 100% voiding have been calculated in the mid-nineties at Pu contents above 12% and isotopic assays Pu240 + Pu242 > 39% (Bloch and Tetart, 1989; Cathalau, 1992). These calculations in general use as an entry point before loading in a MOX PWR, plutonium isotopic compositions of reprocessed UOX with average burnup 33– 45 GWd/t shown in line 1 and 2 of Table 1. They also use conservative calculation hypotheses adapted to these plutonium compositions customary of this time frame in the French fleet. The authors quoted above (Bloch and Tetart, 1989) propose a theoretical laser purification process eliminating 100% of the isotope Pu242 contained in the plutonium of MOX fuel assemblies unloaded after the first recycle to avoid positive dq’s. They also show that elimination of Pu242 is more favorable for reactivity considerations than Pu240 since this latter isotope is converted in fissile Pu241 during the cycle. This important point will be reconfirmed many years later by B. Gastaldi, J. Tommasi and A. Zaetta (Gastaldi et al., 2017) using an analytical model. Rather independently, the French Special Isotope Separation (SIS) team in CEA/Saclay performs an experiment on the same isotope Pu242, on November 6, 1991. This complete separation experiment establishes at a small scale the scientific feasibility of AVLIS separation of isotope Pu242. A comparable feasibility is mentioned in (Alternative Applications of Atomic Vapor Laser Isotope Separation Technology, 1991) issued by the American Academy of Sciences. A recent CEA internal report mentions theoretical extrapolations of this scientific feasibility experiment to isotopic plants, depleting the plutonium content in Pu242 to values allowing multirecycling in PWR’s (Bréchet, 2017). The purpose of this paper

is to give quantitative results of plutonium multiple recycling in a long term fleet consisting of PWR’s of the EPR type with UOX and MOX fuel. To this reactor fleet are added the following fuel cycle facilities: classical PUREX reprocessing and standard MOXMIMAS fabrication plants and theoretical laser isotope depletion plants, extrapolated from the November 1991 scientific experiment. This paper is organized as follows: Section 2 recalls the known limits of Pu recycling in a thermal spectrum and the Water Void Reactivity Effect (WVRE). Section 3 describes multirecycling in a long term EPR and PWR’s 1600 fleet supplemented with fuel cycle plants both classical and innovative. The Tirelire-Strategie cycle code (Massara et al., 2005) is used to simulate the long term operation of the nuclear fleet and the associated fuel cycle facilities. Section 4 discusses the results obtained. Section 5 benchmarks multi recycling in PWR’s 1600 with plutonium AVLIS with a reference scenario where 60 GW of Sodium Fast Reactors operate in France beyond 2080. Section 6 sketches an extension of the stable nuclear capacity scenario at 55, 6 GW to a reduced nuclear fleet where only 16 EPR’s and PWR’s 1600 operate in France beyond 2050. 2. Limits to Pu multi-recycling in a thermal spectrum As outlined in the introduction, water voiding induces a spectrum hardening in a standard PWR. It can be shown that beyond 80% of void in the coolant the neutron spectrum is flattened in the region 0,02–100 eV and shifted to the right, beyond the giant absorption resonances of Pu242 and of Pu240 respectively at 2,7 eV and 1 eV. The absence of resonance absorption of the neutrons then leads to a reactivity increase dq which, when positive has to be avoided for safety considerations. We follow Bruna et al. (1998) to write at least formally the void reactivity effect as follows:

dqvoid ðtÞ ¼ dqmod ðt Þ þ dqboron ðtÞ þ dqcore ðtÞ

ð1Þ

where: dqvoid(t) is the total reactivity worth induced by water voiding dqmod(t) is the reactivity worth induced by the neutron spectral shift. The giant resonance absorptions peaks mentioned in Section 1 are the main physical phenomenon underlying dqmod(t) dqboron(t) is the reactivity worth induced by soluble boron withdrawal with the voiding of the coolant dqcore(t) is the reactivity worth due to core behavior: power distribution, temperature, fission product poisoning, In order to ensure reactor safety one has to satisfy at all times and for all transients

dqvoid ðtÞ þ =e= < 0

ð2Þ

e includes the various errors induced by cross sections and computational tools. In practice dqmod(t) is the void reactivity worth component sensitive to the plutonium content and its isotopic composition. It is this component upon which the fuel designer can act to keep the

Table 1 UOX plutonium isotope compositions from UOX irradiated fuel, before loading in a MOX PWR. Cooling time before reprocessing (for Pu241 decay) is taken as 5 years unless mentioned otherwise. Plutonium in line 3 is second generation Pu from line 1 reloaded in a PWR. Plutonium in line 4 is an isotopic composition averaged on the fleet in 2035 used in fuel cycle scenarios (see definition below). Pu Isotope in at/%

Pu238

Pu239

Pu240

Pu241

Pu242

1-from UOX 33 GWd/t (ref. Cathalau, 1992) 2-from UOX 45 GWd/t (ref. Gastaldi and Guyot, 2015) 3-2nd generation Pu from UOX 33 GWd/t (ref. Cathalau, 1992) 4-average Pu_2035

1,85% 2,85% 2,74% 3,60%

58,05% 55,72% 42,51% 41,20%

22,05% 27,52% 29,19% 35,50%

10,75% 4,37% 14,30% 7,40%

5,60% 9,10% 9,82% 11,60%

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void reactivity worth negative by depleting the Pu242 assay with an AVLIS laser process. The 2 other components dqboron(t) and dqcore(t) have generally been treated since the early nineties in a conservative way to ensure conditions (2) in the worse conditions that can be envisioned. Bruna et al. (1998) outline qualitatively the following limitations when estimating these two components: dqcore(t) incorporates in the neutron balance (or in the reactivity) complex phenomena such as fuel dilatation, moderator density variations, Doppler widening of heavy nuclides resonances, fission products poisoning. dqcore(t) does not depend to first order on fuel composition (in our case plutonium content), but on the core parameters when the voiding transient is initiated. Using classical perturbation theory (Reuss, 2003); it is possible to show that

dqboron ¼ a1  rboron  dNboron þ a1  Nboron  drboron a a

ð3Þ

a1 is a negative constant related to the Boltzmann operator H averaged over the core, therefore when dNboron is negative with water voiding, there is a reactivity increase due to the loss of soluble boron and a positive contribution to dqboron. rboron , the neua tron absorption varies as (En)1/2 therefore shifting of the neutron spectrum to the right (dEn > 0) induce a loss of absorption, a negative drboron in term 2 of Eq. (3) and a supplemental positive a contribution to dqboron. The first term of Eq. (3) is usually quoted as the most important in qualitative analyses. A practical rule which we will follow in Section 3 in order to induce conservatism in respecting Eq. (2) is to maximize dNboron and perform calculations at high boron (Nboron) concentrations, at the beginning of the reactor cycle. These limitations expressed in a qualitative way have been studied and resolved at least partially 18 years after the original analyses by B. Gastaldi and co-workers (Gastaldi and Guyot, 2015). They have used fuel assemblies transport calculations with and without leakage effects and the Apollo 2 code to compute void reactivity worths dqvoid in a number of carefully defined configurations. They have also performed full core computations re-using Rf and Ra averaged over fuel assemblies to take into account boron concentration (related to dqboron), burn-up (related to dqcore), loading patterns (related to dqcore) and leakage effects (related to dqcore). Both 100% MOX cores and 30% MOX cores have been analyzed for homogeneous voiding (void reactivity worth of the complete core) and heterogeneous voiding (void reactivity worth of a fraction of core). A variety of Pu isotopic compositions have been treated both with low even isotopes (Pu242 and Pu240) content and higher content such as in Table 3, input column. Through these detailed analyses, Gastaldi and Guyot (2015) and Gastaldi, 2016 broadly confirm in a quantitative, more precise way the qualitative tendencies outlined much earlier by Bruna et al. and the early reference (Bloch and Tetart, 1989) calculations. Table 2 from recent reference (Gastaldi, 2016) gives computed values of dqvoid in conservative conditions with kinfinite and various boron concentrations Cb (in ppm). The plutonium isotopic vector is the one obtained by averaging the vectors of unloaded MOX fuel assemblies in water storage in 2035, assuming the French single recycling policy is pursued until these years. We label this plutonium ‘‘Averaged_2035”. Its even isotopes content is close to the one of Parity MOX fuel used in the EDF fleet between 2014 and 2019. With such an isotopic quality, dqvoid becomes positive when the plutonium content CPu is between 12% and 13% depending on the boron concentration. The Table 2 results stay for MOX with depleted uranium support (0.25% 235U). There still remains a debate about bounding conditions and robust ways to satisfy condition (2). As was the case in the early nineties papers (Bloch and Tetart, 1989; Cathalau, 1992), full homogeneous voiding of a MOX fuel assembly in a 100% MOX core

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is considered as a bounding case for supposedly more favorable configurations such as voiding in a 30% MOX core. Gastaldi (2016) shows that in homogeneous voiding for 30% MOX cores, neighbouring empty UOX assemblies can have a stabilizing effect on the spectrum shift and lead to dqvoid <0, all other conditions such as plutonium content and Pu isotopic vector remaining as in the 100% MOX core. But the demonstration so far has not been extended to heterogeneous voiding, for instance when only the center part of the MOX assembly is empty and the UOX assembly nearby is full. Since the objective of this paper is not to resolve the reactor physics debates of voiding reactivity worth in MOX cores, both 100% and 30%, we adopt the following simplified approach even if it leads to conservative values of plutonium content at significant Pu242 and Pu240 assays and makes laser isotope depletion even more necessary for multirecycling. Simplified assumption 1: Void reactivity worth numerical calculations are performed with the Apollo 2 code for fuel assemblies representing a 100% MOX core with isotopic compositions (or vectors) coming out of the fuel cycle code (see Sections 3.1 and 3.2). dqvoid is defined with 100% of the water coolant missing from the fuel assembly and is computed as:

dqvoid ¼ qempty  qfull equivalent to dqvoid   empty full ¼ log k =k

ð4Þ

since k, the multiplication factor is very close to 1. Simplified assumption 2: k, the multiplication factor can be computed without leakage in an infinite array of similar fuel assemblies (kinfinite) or with a leakage term representing a finite medium in the form:

  Keff ¼ kinfinite = 1 þ B2g M2

ð5Þ

With standard neutronics notations, Bg the buckling (or geometric laplacian) and M2 the migration area as given by the Apollo transport code. To maximize conservatism we choose to compute dqvoid with kinfinite, without any leakage, though some authors (Gastaldi and Guyot, 2015) have used a simplified Bg value for a cylindrical reactor. Simplified assumption 3: For all these kinfinite calculations, the boron concentration is taken constant at 2000 ppm, a maximizing value equal to the initial value of the boron concentration for an hypothetical 18 months reload of a PWR 1600 with 100% Mox core. The conservatism of simplified assumption 3 is discussed in Section 4, in a practical case. Simplified assumption 4: Finally we perform kinfinite and void calculations at the beginning of the cycle, just after the xenon equilibrium is reached (like boron, xenon has a large cross section in a thermal spectrum. As water is voided, the neutron capture in xenon vanishes and the reactivity rises). To first order, as was assumed in the early nineties, the reduction of the plutonium content with fissions and fuel burnup increase will reduce the void reactivity effect. Again, corrections to this early assumption have been brought by Gastaldi and coworkers (Gastaldi and Guyot, 2015; Gastaldi, 2016). But it ensures conservatism in the dqvoid calculations. It will always be possible to use less conservative hypotheses in future work. For instance, perform fuel assembly calculations with leakage effect. It is then possible to show that the plutonium content CPu at which dqvoid becomes positive is increased by a value around 1%. This would modify in a slightly more favorable way plutonium fluxes and balances in the fuel cycle calculations presented in Section 3. But it would not change its major conclusions

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Table 2 Reactivity void worth dqvoid in pcm at 100% voiding, for various boron concentrations Cb. From Gastaldi (2016). Pu content CPu

Cb = 0 ppm

Cb = 300 ppm

Cb = 600 ppm

Cb = 900 ppm

Cb = 1200 ppm

Cb = 1500 ppm

7% 8% 9% 10% 11% 12% 13% 14% 15% 16%

32,920 26,144 20,175 14,879 10,150 5904 2072 1401 4560 7441

31,788 25,117 19,232 14,005 9334 5138 1349 2087 5212 8062

30,679 24,108 18,304 13,144 8530 4382 636 2762 5854 8676

29,590 23,117 17,392 12,298 7739 3639 66 3429 6489 9281

28,520 22,142 16,495 11,464 6959 2905 759 4086 7114 9879

27,471 21,184 15,611 10,643 6191 2183 1442 4734 7731 10,468

about the feasibility of multi-recycling MOX fuel in PWR’s using plutonium AVLIS to deplete Pu242. We define an ‘‘Averaged_2035” plutonium isotopic vector as the one obtained in the French fleet when averaging over the MOX assemblies unloaded for interim storage with the assumption that the policy on only one plutonium recycling –monorecyclage in French- is pursued until 2035 (and later). We use depleted uranium Udep for the uranium component of the mixed oxide. We now apply the dqvoid methodologies outlined above to compute a reload of such MOX fuel in a 100% MOX PWR with Power level 1600 MWe. The reason of such a choice will be detailed in further sections. We next calculate the Pu content of the reload by keeping an equivalent number of fissile nucleides Pu239 + Pu241 with those of an actual first recycle reload in order to conserve reactor performances (Reuss, 2003). If n (respectively n + 1) denotes cycle (or generation) n (with duration for instance 3  18 months for the 45 GWd/t average burnup at discharge shown in Fig. 3 for PWR’s 1600), then in order to maintain reactor performance over the cycle we have to keep:

 
nþ1 n nþ1 kinfinite Cnþ1 ¼ kinfinite CnPu ; VnPu Pu ; VPu

ð6aÞ

It can be shown that the Pu content CPu increases above the first generation MOX reload value of 8,5% (Parity MOX management) and is given by

10% < C Pu < 13%

ð6bÞ void

We next look up dq (CPu,VPu) in Table 2 using data computed by Gastaldi and co-workers. Using the conservative methods recalled above and high boron concentrations, dqvoid becomes positive and thus inacceptable. One way to overcome this limitation is to modify its isotopic Vector VPu (or composition) and deplete Pu242 when loading the PWR’s 1600: - to values compatible with a negative dqvoid computed with conservative conditions, - to values compatible with an AVLIS plutonium process, - and to fuel cycle conditions since in practice fleet scenarios have to manage 3 plutonium fluxes: the current one from reactors loaded with UOX fuel, the current one from reactors loaded with 100% MOX fuel assemblies and the backlog one of former 900 MW units, in intermediate storage (see Fig. 2 and Section 3.2). Those 3 fluxes have different isotopic vectors VPu, this is why we have introduced the notation dqvoid (CPu,VPu) leading to tabulations of dqvoid for these 2 parameters while keeping the principles outlined above. If one wishes to study uranium effects, the definition of dqvoid would have to be extended to the uranium component and become dqvoid (CPu,VPu, CU, VU). As pointed out above, this is

beyond the scope of this paper, focused on the plutonium component of PWR fuel. 3. Multi-recycling in a long term PWR’s fleet including fuel cycle plants, both classical and innovative 3.1. An idealized isotopic transformation We consider an idealized isotopic separation of Pu242 as given in Table 3. The input of the transformation (or in fact of the isotopic plant) is plutonium coming out of a MOX fuel assembly unloaded from the EDF nuclear fleet around 2015. It is characterized by a high Pu242 content: between 11% and 12%. As explained in Section 2 if such a series of fuel assemblies were loaded in an EPR reactor, with CPu increased to values given by Eq. (6a,b) to have enough fissile nuclides and keep reactor performances, it could result in a positive water void reactivity effect dqvoid after a few months in the reload cycle, when the assay of even Pu isotopes has risen due to irradiation. The reload of MOX fuel assemblies with such an isotopic vector is thus precluded by core safety analysis. Table 3 gives the output of the isotopic transformation. Of the order of 80% of the initial Pu242 is separated in the tails, leaving only 20% of the original Pu242 in the product. It is possible to approximate the isotopic transformation by a 10% increase of the assay in isotopes 238, 239, 240, 241 in the product. The isotopic tail is a mixture of the 5 plutonium isotopes. Its mass output is of the order of 15–20% of the transformation input. 3.2. Application of this transformation to the stable nuclear scenario beyond 2050 Fig. 2 shows the application of this transformation to a 55,6 GWe nuclear fleet consisting in the 2053 time frame of: - 4N4 units (1500 MWe) loaded with UOX fuel. When these units reach their end of operation, they are replaced by an equivalent capacity of EPR NM reactors; - the FOAK EPR Unit at Flamanville; - 20 EPR-NM reactors loaded with UOX fuel;

Table 3 An idealized laser isotope transformation (Camarcat and Lemaire, 2016). For one kg of plutonium feed, the isotopic transformation yields 825 g of product recycled in the PWR reactors and 175 g of tails conditioned as waste in a ceramic matrix. Pu isotope

Transformation input (or plant Input) (French Parity MOX fuel assembly at unloading 2015)

Transformation output (or plant output)

238 239 240 241 242

3,43% 39,31% 29,63% 16,50% 11,13%

3,75% 43,05% 32,45% 18,07% 2,67%

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- 10 1600 MWe PWR’s loaded with MOX fuel, 100% of the core consisting of MOX fuel; - classical fuel cycle plants for reprocessing at La Hague and for MOX fuel fabrication at MELOX site in Marcoule. For clarity, these plants are not shown in Fig. 2. In order to deplete the Pu242 content in plutonium before entering the 100% MOX PWR’s for recycling, two isotopic plants are added to the classical reprocessing and MOX fabrication fuel cycle facilities. In the upper part of Fig. 2, plant EU420 depletes Pu242 from the backlog of PWR MOX assemblies which have been stored in water pools after being unloaded from the 900 MWe units. Assuming that the French nuclear fleet continues on the actual trend until 2050, this backlog stock will reach 5000 metric tons of MOX fuel at this time containing 265 tons of Pu. In the lower part of Fig. 2, we add a second isotopic plant EU42M to deplete Pu242 from the current plutonium coming out of the MOX 1600 PWR’s and allow its recycling. This current plutonium is mixed with the plutonium coming of the EPR reactors loaded with UOX fuel, whose Pu242 content is lower than in the input column of Table 3. The two isotopic plants correspond to two different stages of the technological development of plutonium AVLIS (in France). The first plant EU 420 uses results of the early nineties (Camarcat, 1994). The second one, EU42M incorporates to the EU 420 design results of the uranium AVLIS large scale experiments performed by French CEA in Pierrelatte in 2003 (Bordier and Lemaire, 1995) Both plants rely on the same isotopic transformation given in Table 3, but their internal design and flux capabilities are different. Examination of Fig. 2 shows that plant EU 42 M redesigned in 2016 is capable of higher plutonium throughputs. In fact it is easier when running the Tirelire-Strategie cycle code to perform the mix of the plutonium fluxes coming out of the isotopic plants of Fig. 2 in a simplified way since the 2 plants have the same isotopic transformation. This is shown in Fig. 3 where a single plant EU(420 + 42 M) is used before loading the two MOX fluxes in the PWR 1600 full MOX reactors. A further isotopic balance is realized with the plutonium of the UOX fuel which does not need depletion of isotope Pu242. Adding such a transformation on plutonium of comparatively low Pu242 assay would be too costly.

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In practice CPu is adjusted by the fuel cycle code to the isotopic vector VPu to keep the performances of the 1600 PWR’s and maintain reactivity through the cycle. The more the even isotopes content, the higher CPu (see Section 2 and Eq. (6a)). Isotopic vectors VPu are computed at important points of Fig. 2. dqvoid is computed outside the cycle code to check that it remains always negative. It is possible to assess the conservatism of the dqvoid (CPu,VPu) calculation by performing Monte Carlo Calculations of the cores for significant values of parameters CPu and VPu. The plants are run from 2050 to beyond 2130. We choose to detail the plutonium fluxes in the year 2070: - First we try to process the backlog of PWR-MOX fuel assemblies in a relatively short period of time, of the order of 25 years in the time frame 2050–2075. To do this, we impose the processing of 200 t of backlog PWR 900 MOX each year; - second we endeavor to design a fleet structure similar to the one used in EDF around 2012, with 22–900 MW units- loaded with MOX fuel (30% in the core) out of a total of 58 units. The balance between standard UOX units and MOX loaded units in Figs. 2 and 3 remains roughly the same: 2/3 UOX and 1/3 MOX. This allows continuity of industrial practices for fuel management, outages placement, fuel fabrication, fuel transport, etc. The usefulness of keeping such an approach in industrial practices is not always apparent to fuel cycle code engineers, in general more focused on their numerical tools. The counterpart of such an approach is that high MOX content reactors are needed to absorb the 2 plutonium fluxes, the current one and the backlog one. For this reason, we use 100% MOX loaded PWR’s with normalized output of 1600 MWe. Those may differ from the EPR-NM basic design which is currently underway in EDF. We keep an identical reactor vessel and an equivalent core: 241 Fuel Assemblies, 17  17 and 14 feet long. This is more advanced than the conventional 30% loading pattern of 900 MWe units and would require industrial development to modify the EPR NM model. Still, for the first 50 years of the scenario 16 GWe with 100% MOX cores can absorb the fluxes both of the backlog and of the current circulating plutonium.

Fig. 2. Schematic plutonium Fluxes – Stable Nuclear Scenario.

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Fig. 3. Mixed reactor fleet during the backlog plutonium consumption phase.

Beyond 2118, it is possible to convert 100% MOX units to standard UOX reload while keeping the same generating capacity. This is shown in Fig. 4. 3.3. The Tirelire-Strategie fuel cycle code Tirelire-Strategie (Massara et al., 2005; Massara et al., 2006) is a calculation code aimed at simulating the operation of a nuclear fleet and the associated fuel cycle facilities over a long period of time (decades, even centuries). It is used to analyze the

consequences of strategic choices related to the nuclear fleet composition (reactors and fuels) and other fuel cycle facilities features. The main reactor types are currently modeled in TirelireStrategie, with models based on CEA reference neutronic codes: current and advanced PWR modelled with APOLLO2 (Sanchez, 2010), several FBR designs and Accelerator-Driven Systems (ADS) – with models based on ERANOS (Rimpault et al., 2002). The validation of Tirelire-Strategie was carried out by the following code-to-code comparisons: APOLLO2 (Sanchez, 2010), for PWR-only scenarios; ERANOS (Rimpault et al., 2002), for FBR-

Fig. 4. Long term equilibrium (2150) multirecycling Pu in 12.3 GWe MOX.

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only scenarios; COSI4 (Grouiller, 1991), the CEA code for fuel cycle studies, on scenarios including both PWR and FBR. Tirelire-Strategie allows nuclear scenarios simulation to comply with industrial requirements (such as spent UOX and MOX reprocessing capacity limitation, interim storage capacity, cooling time before reprocessing, delay for fresh fuel fabrication, losses during reprocessing or fabrication, number and characteristics of reactors being reloaded for each year) and to take into account strategic choices (i.e. type of reactors and fuel management used for the nuclear fleet renewal, minor actinides incineration rates, interim storage management) (Laugier et al., 2005). The main parameters defining the dynamics, year per year, of a nuclear scenario in Tirelire-Strategie are basically:  The nuclear fleet installed capacity (in GWe), which is related to the electricity production via the average fleet load factor;  The installed capacity of each nuclear system type (i.e. PWR UOX, PWR MOX, Na-cooled FBR, etc.);  A priority level associated to the deployment of each nuclear system type;  Maximum introduction rate for the total fleet and for each reactor type;  MA (Np, Am and Cm) rate at fuel fabrication and losses at reprocessing for all actinides and for each reactor type;  Reprocessing rate for each fuel type;  Spent Nuclear Fuel (SNF) cooling time before reprocessing and delay for fresh fuel fabrication, for each reactor type. On the other side, each reactor type is characterized by:  Maximum lifetime;  Core Heavy Metals (HM) mass and the HM mass reload (taking into account its reload batch size);  Fuel type which can be charged in each reactor type (i.e. UOX or MOX fuel in PWRs, MOX fuel for FBRs, Pu on The support for advanced PWRs);  Fuel irradiation time;  Parameters specifying the models for the calculation of the fuel irradiation (evolution model) and the Pu content for MOX fuel (equivalence model). The evolution and equivalence models are different for PWR and FBR (Massara et al., 2005). The code calculates the power capacity to be installed every year, on the basis of the total power demand and the number of decommissioned units, if any. This power demand will be satisfied

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by the reactor types taken into account in the current scenario, following their priority level and their maximum introduction rate: i.e. if the priority is 1 for PWR, 2 for Na-cooled FBR and 3 for HTR, then PWR will be deployed up to their maximum deployable power. When this limit is reached, the next reactor type in the priority order will be considered (in this case FBRs) and so on. Once the nuclear fleet composition determined, the discharged fuel composition is assessed by means of the evolution models described in (Massara et al., 2005). After the Spent Nuclear Fuel (SNF) cooling simulation, the reprocessing is modelled considering:  The corresponding rates for each fuel type;  The fuel management strategy for each reactor type. As a result of the SNF reprocessing, Fission Products (FP) are sent to geological disposal in vitrified glass logs, whereas Pu and MA, depending on the fuel management strategy, either are sent to geological final repository, or feed the correspondent in-cycle stocks for further in-cycle handling. 4. Results and discussion 4.1. Reprocessing and fuel fabrication capabilities Referring to Figs. 3 and 4, we choose to prioritize reprocessing of backlog MOX fuel assemblies, starting in 2050. Fig. 5 shows reprocessing capabilities versus time for the 3 types of fuel: EPRUOX, PWR 900 MOX (backlog) and PWR 1600 MOX (current). The backlog PWR 900 MOX fuel is reprocessed at a capacity of 200 t/yr, allowing pool clean-up between 2050 and 2073 (brown curve). Reprocessing of PWR 1600 MOX assemblies starts in 2055 at 100 t/yr, increases to 200 t/yr in 2065 and 400 t/yr in 2073 when the backlog PWR 900 MOX fuel has been dealt with. In practice we try to keep a maximum total MOX reprocessing flux (i.e backlog PWR 900 MOX + current PWR 1600 MOX) at 400 t/yr. This lasts until around 2080. Then reprocessing of PWR 1600 MOX decreases with the total GW capacity of PWR’s 1600. In 2138 a long term equilibrium is reached with total MOX fuel reprocessing flux of 231 t/yr and total UOX fuel reprocessing of 812 t/yr. Fig. 5 shows that the long term reprocessing equilibrium beyond 2150 is similar to the one achieved at La Hague around 2010–2015: 1043 t/yr of reprocessed fuel. However, its composition is markedly different since only UOX fuel is reprocessed at La Hague on an industrial basis in the 2010–2016 time frame. In

Fig. 5. Reprocessing capabilities during the scenario (t/yr). PWR MOX (brown curve) are the backlog assemblies unloaded from the 900 MW units. The total red MOX curve consists of the 2 MOX fluxes: PWR 900 (backlog) and PWR 1600 (current). (For interpretation of the references to colour in this figure legend, the reader is referred to the web version of this article.)

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Fig. 6. Water storage of UOX and MOX (tons of Heavy Metal) fuel assemblies versus time. PWR MOX (brown curve) are the backlog assemblies unloaded from the 900 MW units. The total red MOX curve consists of the 2 MOX fluxes: PWR 900 (backlog) and PWR 1600 (current). (For interpretation of the references to colour in this figure legend, the reader is referred to the web version of this article.)

such a PWR’s 1600 scenario, industrial chemical processes would have to be developed to treat separately 2 incoming streams: standard UOX fuel assemblies and standard MOX ones (current and backlog). Similar curves can be drawn for the fuel fabrication capabilities. Maximum MOX fabrication capacity reaches 300 t/yr between 2060 and 2100 when the 10 PWR’s 1600 are run to load both backlog PWR 900 plutonium and the current one, recycled from UOX EPR’s and 100% MOX PWR’s 1600 (see Fig. 3). In 2131, a long term equilibrium is reached for fabrication. It is similar to the long term equilibrium for reprocessing. 231 t/yr of MOX fuel fabrication and 812 t/yr of UOX fuel for the 2 types of reactors of Fig. 4. 4.2. Intermediate storage in pools The current single recycle scenario of the French fleet is run from 2015 till 2050. The backlog of PWR 900 MOX fuel assemblies reaches 4500 tons cumulated around 2050. Then the isotopic depletion plants EU420 and EU42M are switched on. Figure shows that the backlog (brown curve) is brought back to zero in less than 25 years, in 2073. This is a strong feature of this multirecycling scheme since accumulation of backlog assemblies without retrieval might be seen as a dead end of the French single recycle strategy. Beyond 2140, the long term part of Fig. 6 is also of interest. The PWR 1600 MOX assemblies in short term water storage (5 years) level down to a cumulated figure of 1150 tons. The EPR UOX fuel assemblies in short term water storage level down to a cumulated value of 4060 tons. The total UOX + MOX storage capacity of 5300 tons is limited to approximately 1/3 of the one existing around 2015. And this is a short term water storage for cooling until reprocessing. In other words, Pu242 depletion with plutonium AVLIS (transformation given in Table 3) gives one solution to MOX fuel accumulation in water pools, including backlog fuel unloaded between 1990 and 2050. It may not be the only way to recycle current and backlog plutonium in thermal spectrum reactors without modifying the industrial design of reactors and of MOX fuel assemblies and their fabrication plants. Another way for instance to overcome the dqvoid limit when CPu reaches values above 12% could be to replace fissile Pu atom by U235 ones. However depletion of Pu242 is a scientific possibility. It is not possible to make an economic comparison between multirecycling with Pu242 depletion and multirecycling with enriched U235 and plutonium oxide new fuels. Preliminary cost figures have been given in the 19900 s literature for plutonium AVLIS plants but it is not possible to relate their performances to those

of the two plants outlined in Fig. 2. What can be said is that the technological readiness level of enriched uranium and plutonium oxide fuel plants seems probably higher than the readiness of Pu242 AVLIS plants. 4.3. Plutonium inventory in the global cycle (reactors, plants, storages) The total plutonium inventory in the reactor cores, in the cycle plants (reprocessing and fabrication), and in the water storage pools is shown in Fig. 7. The linear increase of plutonium in water storage (i.e 7 to 8 t/yr between 2005 and 2050) is stopped when MOX reprocessing starts accompanied by Pu242 depletion in the isotopic plants. In the long term around 2150, the total plutonium inventory is stabilized at 370 t, after a maximum value in the transient at 640 tons, around 2075. The total plutonium inventory is generally used to benchmark multirecycling scenarios in fast reactors, in thermal and epithermal ones. A low inventory is considered as a premium by industry in such scenarios, usually for economic reasons since transporting a given throughput of plutonium at the industrial scale is more costly than the equivalent uranium flux. To this low inventory it is mandatory to add the plutonium tails which may either be considered as an ultimate waste or may be conditioned in retrievable waste forms for burning in fast reactors in the 23rd century. We first study the least favorable of these 2 options, and suppose that the tails of the 2 isotopic plants are categorized as ultimate waste, to be disposed of. 4.4. Isotopic tails and actinide (major and minor) wastes Figure shows the 4 principal actinide waste streams coming out of the reprocessing and isotopic plants. Of course classical fuel cycle plants also produce fission products waste (vitrified in glass logs) and transuranic waste packaged according to the industrial standards of reprocessing and MOX fabrication. Those ‘‘classical” streams are not shown in Fig. 8. For reprocessing plants, we discuss first the minor actinide fluxes: Am, Np, Cm. We underscore the importance of the americium flux, unavoidable between 2050 and 2075 due to the processing of old MOX fuel assemblies including plutonium whose isotope Pu241 has decayed into Am241 (halflife of 14 years). The tails of the EU42 isotopic plants at high Pu242 assay reach a peak of 4 t/yr around 2115 and level down afterwards to an asymptotic value of 3 t/yr in 2140. Such a high value is unusual in French fuel cycle calculations (and real life) with single plutonium recycling. But one should not forget that

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Fig. 7. Plutonium inventory in the global cycle (tons).

Fig. 8. The 4 categories of waste from the isotopic plants (plutonium tails) and reprocessing plants (minor actinides).

such fluxes will be buried in geological disposals of European northern countries in the time frame 2022–2026. Sweden and Finland operate altogether of the order of 15 reactors with an open cycle. Those reactors produce at fuel unloading 2,5–3 t/yr of plutonium in ceramic UO2 matrices of the fuel assemblies. These matrices are themselves enclosed in large copper casks, with ultrahigh containment capabilities. If waste disposal of plutonium tails is chosen and since it would be a single chemical species, we recommend moving away from the vitrified industrial glass logs well suited to multi-species packaging. plutonium could be conditioned in a ceramic matrix, allowing a higher incorporation ration of plutonium in the waste package. This kind of research has been performed by CEA at the laboratory scale in the late nineties for possible conditioning of separated minor actinides. Provided that a 10–15% plutonium incorporation ratio is obtained in a ceramic process at the industrial scale, packaging a flux of 3 t/yr would require only of the order of 100 ceramic logs per year. This is an order of magnitude below the standard glass logs flux produced at La Hague when reprocessing 1050 tons per year of fuel as shown around 2140 in Fig. 5 and in actual industrial life in 2015. Therefore, if R&D is successful, the disposal of well-conditioned plutonium isotopic tails would not overburden the French CIGEO waste disposal with respect to standard reprocessing. Details are given below: Evaluation of the disposal area of these waste packages and comparison with the disposal area of MOX fuel in the singlerecycling scenario.

We evaluate the disposal area of the waste packages containing 30 kg of plutonium in a ceramic matrix. We compare the disposal areas when all the wastes produced in the 2 scenarios have been disposed of underground, and no wastes –either isotopic tails or MOX fuel assemblies in pools- are left above the ground. We perform the area comparison after a decay time of 90 years which is necessary for Fission Products (FP’s) decay in MOX fuel assemblies. Pu242 tails ceramic logs could be sent to underground disposal in an earlier time frame since they do not have Fission Products nor short half life minor actinides like Cm244 incorporated. The tails packages contain a majority of Pu242 (between 40% and 50%). This isotope has a very low thermal cross section and does not fission in a thermal spectrum. It implies that nuclear criticality is not a concern from a repository perspective. Pu multirecycling with AVLIS leads to tails with isotopes Pu238 and Pu241. They cause significant decay power around 100 years, the time frame of comparison with MOX fuel assemblies in casks after the decay of their Fission Products. The decay heat of a MOX fuel cask containing one ton of heavy metal is 2310 W after a cooling period of 90 years in order to wait for the decay of Fission Products. At the same time, the decay heat of a ceramic package containing 30 kg of Pu242 tails is only 330 W. The ratio between those 2 decay heats is 0,14. To first order the disposal area per package is proportional to the decay heat, This hypothesis has been checked in the case of the clay disposal project in France. Thus the disposal area of a Pu242 tails waste package is considered as 0.14 times the one of a one ton MOX assembly. The cumulated disposal areas of the 2 scenarios are shown in Table 4. The yearly isotopic tails fluxes are obtained from the AVLIS multirecycling scenario in Figs. 2–4. They are linearly converted to Pu242 tails packages per year in the second column of Table 4. We make only relative comparisons with MOX fuel assemblies disposal. We compute the disposal area in a unit defined only for comparison purposes: equivalent MOX tons. At the end of the single recycling scenario, 11,400 tons of MOX fuels have to be disposed of. At the same time, the isotopic tails packages would use only 1560 of the equivalent area units. In this sense, AVLIS multirecycling is 7 times more efficient than singlerecycling in terms of disposal area. This is a significant quantitative gain. Also the Pu242 tails packages could be brought down a few years after their casting while the MOX fuel casks have to wait Fission Products decay for about 90 years in intermediate storage. Another important aspect of disposal is the easier handling capabilities of plutonium ceramic logs. Assuming as above a conservative 10–15% mass incorporation ratio of plutonium in the ceramic matrix, the Pu242 waste packages would have a total mass of a few hundreds of kilograms, including the metallic container and a radiation shield. This would be an order of magnitude below

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Table 4 Comparison of cumulated disposal areas in the 2 scenarios. The AVLIS multirecycling scenario is 7 times more efficient than the single-recycling one. Multirecycling AVLIS scenario Pu 2050 2060 2070 2080 2090 2100 2110 2120 2130

242

tails Packages/year

734 109 146 142 148 153 130 130 117

Single-recycling MOX scenario Area per year (equivalent MOX tons)

Cumulated area (equivalent MOX tons)

MOX tons for disposal

11 16 21 20 21 22 19 19 17

0 158 369 575 789 1012 1201 1390 1560

5000 5800 6600 7400 8200 9000 9800 10,600 11,400

the weight of a MOX assembly large fuel cask (several tons). In order to complete this section on the capabilities of ceramics to package plutonium waste, we point out that R&D centers on pyrochemical processes also perform research on ceramic immobilization of salt wastes containinig transuranic long lived elements.

4.5. Plutonium isotopic composition at the input of the PWR’s 1600 The Tirelire-Strategie cycle code self consistently computes at various times the plutonium isotopic composition of the mix resulting of the 3 fluxes shown in Fig. 2: - the depleted plutonium coming out of the PWR 900 backlog MOX fuel assemblies, - the depleted plutonium reprocessed from the current PWR’s 1600 MOX fuel assemblies - the non-depleted plutonium reprocessed from the UOX EPR fuel assemblies. This plutonium mixture is then used to load the PWR’s 1600 with a plutonium content CPu calculated in Tirelire-Strategie to have enough reactivity to reach the fuel burnup of 45 GWd/t. It is therefore possible to plot the isotopic vectors of this mixture with respect to time and using the principles of Section 2 to calculate dqvoid (CPu,VPu) with our 4 conservative assumptions. The scenario is acceptable only if dqvoid stays negative. The time evolution of VPu is given in Table 5a. dqvoid is plotted against time in Fig. 9. With 2000 ppm boron concentration, the value at the beginning of the cycle, the Integral void worth dqvoid is always negative, as shown in Fig. 9. It reaches a peak value 1626 pcm in 2105 when the Pu content is maximum (11.5%). Supposing that we add a penalization of 500 ppm to the boron concentration, increasing it to 2500 ppm, we would still obtain a peak negative dqvoid at 586 pcm. Such a margin with penalization is deemed sufficient to undergo a thorough safety analysis, if necessary supported by full core calculations.

4.6. Natural uranium used Today, in the French feet, 10% of the fuel assemblies are MOX fuel, which leads to a 10% economy of the natural uranium utilised. At equilibrium (Fig. 4) of our scenario, 22% of the fleet uses MOX fuel, and so the natural uranium economy is improved by the same factor. We give in Table 5b below the uranium quantities used by our scenario: 17.2 tUnat/TWhe. We give also the uranium that would be used in a open cycle fleet, with the same reactors characteristics. 5. Benchmarking with Sodium fast reactor multirecycling The CEA report in reference (Bréchet, 2017) has proposed to compare various light water reactor multirecycling technologies with a reference scenario where Sodium Fast Reactors (SFR’s) have replaced around 60 GWe of light water reactors in the French nuclear fleet around 2080. A number of papers have been published at conferences for this SFR reference scenario, see for instance (Camarcat et al., 2011) for homogeneous cores of the early 20100 s and references therein. The main technical criteria and quantitative parameters for the benchmark with the reference SFR’s scenario are the following: - total plutonium inventory in the whole cycle, including reactors, intermediate wet storage, all fuel cycle plant whether classical or innovative, - plutonium inventory in waste, - other transuranic inventory in waste, - use of available uranium resources. The results of this benchmarking analysis are gathered in Table 6. The total plutonium inventory has already been dealt with in Section 4.3. A higher plutonium inventory was reported in ref. (Bréchet, 2017), appendix 7: 480 tons at stabilization in 2150 instead of 370 tons in Fig. 7. This is due to the fact that those very

Table 5a Isotopic plutonium composition VPu during the multirecycle scenario. Pu content (%) Pu even Pu odd Integral Void effect Cb = 2000 ppm Pu238 Pu239 Pu240 Pu241 Pu242

2031 8,7% 34,9% 65,1% 15,656 2,6% 56,4% 25,1% 8,7% 7,2%

2040 8,6% 34,7% 65,3% 16,187 3,0% 57,5% 24,9% 7,8% 6,9%

2050 8,4% 34,1% 65,9% 17,321 2,7% 57,8% 24,7% 8,1% 6,7%

2060 9,8% 39,0% 61,0% 9198 2,9% 54,3% 31,3% 6,6% 4,8%

2070 10,3% 40,4% 59,6% 6919 3,0% 53,0% 33,2% 6,6% 4,3%

2080 10,8% 42,4% 57,6% 4557 3,6% 48,5% 35,6% 9,1% 3,2%

2090 10,9% 42,9% 57,1% 3877 3,8% 47,3% 35,9% 9,9% 3,2%

2105 11,5% 44,1% 55,9% 1626 4,7% 44,4% 36,7% 11,5% 2,8%

2110 11,4% 44,0% 56,0% 2021 4,9% 44,2% 36,2% 11,8% 2,9%

2130 11,2% 43,9% 56,1% 2676 5,6% 43,7% 35,3% 12,4% 3,0%

2150 11,0% 42,9% 57,1% 3441 5,6% 45,5% 33,9% 11,6% 3,4%

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Fig. 9. dqvoid (CPu,VPu) with VPu given in Table 5a and CPu calculated from Eq. (6a) as programmed in the Tirelire-Strategie fuel cycle code.

Table 5b Natural uranium consumption in tons/TWhe for the 3 scenarios under comparison.

Fraction of power produced by MOX Uranium consumption tons Unat/TWhe

Open cycle

Single MOX

Multirecycling with AVLIS

0%

10%

22%

21,0

18,9

17,2

early multirecycling calculations with the techniques outlined in Sections 2–4, used fuel management with burn up of 50 GWd/t instead of the value of 45 GWd/t in this publication. Going to a lower burn up allows lower plutonium concentrations CPu’s at equivalent times and an overall reduced plutonium inventory. Two streams constitute the plutonium inventory in waste: The first one is made up of the isotopic tails of the 2 depletion plants EU 420 and EU 42 M. It has been dealt with in Section 4.4, either as final disposal packages or in retrievable form for delayed transmutation in a fast neutron spectrum. The second stream is the plutonium small flux not retrieved in the standard PUREX process. It amounts to 0,1% (103) of the plutonium throughput in the La Hague plutonium shops. Fig. 5 shows a reprocessing throughput of 1043 t/yr of used MOX and UOX fuel beyond 2130. The equivalent plutonium throughput is of the order of 26,5 t/yr, yielding a Pu waste stream to be vitrified in glass logs I’0 = 265 kg/yr. This is higher but still within the order of magnitude of the equivalent Pu waste stream I0 in the reference SFR scenario. Reference

(Camarcat et al., 2011) gives an SFR MOX flux of 400 t/yr reprocessed in 2150. Assuming a standard PuO2 content of 15% (homogeneous core) when unloading and 60 t/yr of reprocessed oxide, the waste stream I0 would reach 60 kg/yr. Between the 2 scenarios there is a difference by a factor 4 due to the reprocessed fluxes but the main difference lies in the isotopic tails of the plutonium AVLIS plants and a waste stream of 7,5 kg Pu/TWh. If intermediate storage of these tails in a retrievable waste form is chosen pending transmutation in delayed fast reactors, for this criterion the 2 scenarios are equivalent. The environmental impact if the SFR reference scenario amounts to 4,3 kg/TWh of Minor Actinides (Boullis, 2017). Extending the single recycling strategy would give a value of 17,5 kg/TWh of Minor Actinides in disposal. The figure for multi recycling with plutonium AVLIS is in between the upper values: 6,9 kg/TWh of Minor Actinide, with the isotopic tails being accounted for separately (see Table 4). This in between value is attributed to the cut-off of higher minor actinides (i.e Cm243 and Am243) since the father nuclide Pu242 is depleted by the isotopic transformations. Such a cut-off of higher actinides due to plutonium depletion on a different isotope is also mentioned in reference (Forsberg, 2015).

6. A PWR multi recycling fleet with a reduced power capacity At the beginning of this research, two reactor scenarios were envisioned. The first one long term stable nuclear capacity at 55,6 GWe has been explored in complete details above. The second

Table 6 Comparison of the homogeneous core SFR’s reference scenario (Camarcat et al., 2011) with Multi-recycling in PWR’s 1600 with plutonium AVLIS for Pu242 depletion. Around 2150 « stabilized material flows »

Nuclear Fleet with 100% SFR’s (around 60 GWe)

Multi-recycling in PWR’s 1600 with plutonium AVLIS for Pu242 depletion (55,6 GWe)

Comments

Pu inventory in the whole cycle (reactors, intermediate water storage, fuel cycle plants). Pu inventory in wastes

1070 t in 2150

370 t in 2150

30% of the SFR inventory at the same time 2150. Elimination of backlog MOX assemblies possible.

I0 of the order of a few tens of kg/ yr in glass logs (103  reprocessed Pu flux) 4,3 kg of Minor Actinides/TWhe

I’0 (see below) for PWR’s + 3 t/yr of isotopic tails

Single chemical species: therefore possibility to use R&D on ceramic conditioning to increase loading in waste package

Environmental Impact: Transuranic wastes and releases. Use of raw nuclear materials

Consumes depleted uranium

6,9 kg Minor Actinifes/ TWhe + 7,5 kg Pu/TWhe (counted above) 17,2 t Unat/TWhe

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Fig. 10. Lower nuclear scenario beyond 2050 (approximate flux values).

one with a reduced nuclear capacity and only 16 EPR’s and PWR’s 1600 beyond 2050 is sketched in Fig. 10. It will be the object of detailed calculations to be reported later.

7. Conclusion We have shown that the original plutonium AVLIS ideas put forward in the early nineties can be quantified with an idealized isotopic transform and assembled through a fuel cycle code to allow plutonium multi – recycling in PWR’s. Lowering the Pu242 assay is one way to overcome the positive dqvoid reactivity worth at high CPu contents which would appear in the scenario as time evolves. The use of a limited number of PWR’s 1600 loaded with 100% MOX is a convenient way to absorb high plutonium fluxes. Such high fluxes arise when one wants to empty spent fuel pools from the backlog MOX assemblies accumulated during the operation of the PWR 900 fleet between 1990 and 2050: of the order of 5000 tons around 2050. This unloading can be accomplished in about 25 years between 2050 and 2075. Plutonium AVLIS for depleting Pu242 may not be the only way to perform plutonium multi recycling. Another way for instance to overcome the dqvoid limit when CPu reaches values above 12% could be to replace fissile Pu atom by U235 ones. The resulting oxide mixture would consist of slightly enriched uranium and recycled plutonium oxide with a lower plutonium content CPu. The reader is referred to the internal CEA report in (Bréchet, 2017) which has listed and analyzed other technologies as well as their advantages and their drawbacks. However depletion of Pu242 is a scientific possibility. Perhaps the most useful results of the research reported here lie in the methods used to analyze the critical points of multi recycling in PWR’s: -the calculation of the void reactivity worth dqvoid and the conservative assumptions for homogeneous fuel assemblies, particularly when the MOX plutonium content CPu lies in the range 8%– 12,5% - the isotopic transformations - the assembly of the various components of the reactor fleet and of the fuel cycle plants, both classical and innovative.

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