Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system

The performance of the fusion–fission hybrid system based on the molten salt (flibe) blanket, driven by a plasma based fusion device, was analyzed by comparing transmutation scenarios of actinides extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT.

Nuclear Engineering and Design 330 (2018) 241–249 Contents lists available at ScienceDirect Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system T R Plukienėa,⁎, A Plukisa, L Juodisa, V Remeikisa, O Šalkauskasa, D Ridikasb, W Gudowskic a Center for Physical Sciences and Technology, Savanoriu 231, LT-02300 Vilnius, Lithuania International Atomic Energy Agency, Vienna International Centre, PO Box 100, 1400 Vienna, Austria c KTH (Royal Institute of Technology), AlbaNova University Centre, 106 91 Stockholm, Sweden b A R T I C L E I N F O A B S T R A C T Keywords: Fusion–fission hybrid system Incineration of trans-uranium elements LWR and RBMK spent nuclear fuel The performance of the fusion–fission hybrid system based on the molten salt (flibe) blanket, driven by a plasma based fusion device, was analyzed by comparing transmutation scenarios of actinides extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT The IAEA nuclear fuel cycle simulation system (NFCSS) has been applied for the estimation of the approximate amount of heavy metals of the spent nuclear fuel in Sweden reactors and the SCALE code package has been used for the determination of the RBMK-1500 spent nuclear fuel composition The total amount of trans-uranium elements has been estimated in both countries by 2015 Major parameters of the hybrid system performance (e.g., kscr, keff, Φn(E), equilibrium conditions, etc.) have been investigated for LWR and RBMK trans-uranium transmutation cases Detailed burn-up calculations with continuous feeding to replenish the incinerated trans-uranium material and partial treatment of fission products were done using the Monteburns (MCNP + ORIGEN) code system About 1.1 tons of spent fuel trans-uranium elements could be burned annually with an output of the GWth fission power, but the equilibrium stage is reached differently depending on the initial trans-uranium composition The radiotoxicity of the remaining LWR and RBMK transmuted waste after the hybrid system operation time has been estimated Introduction When trans-uranium elements (TRU) are removed from the discharged fuel destined for disposal, the radiotoxic nature of the remaining materials drops below that of natural uranium ore within a period of 500 years (Ewing, 1999) Hence the possibilities of partitioning and transmutation of the long-lived radioactive waste into stable or short-lived isotopes, which could then be surface-stored with the little/no proliferation value, are now under investigation In addition, TRU elements could serve as fuel for transmutation systems The transmutation steps include the reprocessing process, fuel fabrication, management of secondary wastes, etc and it is likely to be the most challenging issue to be solved in the sustainable nuclear fuel cycle These problems are comparable for all types of transmutation systems since they are related to the specific TRU content of the transmutation dedicated fuels (Salvatores, 2009; Yurov and Prikhod’ko, 2014) The idea of the “closed” fuel cycle of plutonium/higher actinides and TRU recycling could be implemented in the future fast neutron systems (Salvatores, 2002) The fast reactor technology is one of the most promising when compared with others at present Some commercially available examples such as BN-600, BN-800 in Russia or FBTR (Fast Breeder Test Reactor) in India could be mentioned Other FBRs (mostly research) have been built and operated in the United States, the United Kingdom, France, the former USSR, India and Japan (Waltar and Reynolds, 1981) It is true that light-water reactors dominate nuclear power today due to nowadays relatively low uranium prices and availability, lower capital costs (by 25%) compared with fast neutron reactors, but still in Generation IV reactor initiative three from six GEN IV reactors are of the Fast Reactor (FR) type: Gas-cooled Fast Reactor (GFR), Sodium-cooled Fast Reactor (SFR), Lead-cooled Fast Reactor (LFR) and Supercritical Water-cooled Reactor (SCWR) (GEN IV, 2016) The latest system has an option for the actinide management based on the particular core design with the fast neutron spectrum (Oka, 2010) The MSFR concept is also recognized as one of Gen IV options because of the compact size and relatively low costs to both build and operate (less metal is needed to fabricate/maintain and no initial fuel Abbreviations: ADS, Accelerator Driven System; Flibe, LiF-BeF2-(HN)F4 molten-salt blanket; FP, Fission Products; LWR, Light Water Reactor; MA, Minor Actinides; SNF, Spent Nuclear Fuel; RBMK, High Power Channel-type Reactor (in Russian abbreviation); TBR, Tritium Breeding Rate; TRU, Trans-Uranium Elements ⁎ Corresponding author E-mail address: rita.plukiene@ftmc.lt (R Plukienė) https://doi.org/10.1016/j.nucengdes.2018.01.046 Received 25 September 2017; Received in revised form 20 January 2018; Accepted 27 January 2018 Available online 20 February 2018 0029-5493/ © 2018 The Authors Published by Elsevier B.V This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/BY-NC-ND/4.0/) Nuclear Engineering and Design 330 (2018) 241–249 R Plukienė et al power, a 250 MW power fusion device operating with a deuterium–tritium fuel cycle is sufficient to provide an external 14 MeV neutron source of × 1019 n/s In (Ridikas et al., 2006) we have shown that in spite of very promising results on the efficient TRU destruction a significant quantity of curium isotopes is accumulated, while for the rest of actinides fully equilibrated concentrations are reached In the parallel paper (Plukiene et al., 2006) an optimization procedure has been developed in order to determine the optimal fission-blanket composition corresponding to the fast incineration rate of all actinides For this purpose a different fissionblanket composition (based on Be + F-salt: 14.29% – 6Li + 7Li, 57.14% – Be, 28.57% – F (Be salt)) corresponding to the fast incineration rate of actinides including minor actinides (also Cm isotopes) was investigated In both F-salt and Be + F-salt cases the same incineration rate (1.1 tons/year) of minor actinides was obtained, but Be + F-salt showed better hybrid-system performance characteristics and a smaller actinide mass in the blanket: kscr at 0.82 ± 0.002 and lower fusion power of 260 MW at equilibrium, which is reached after about 700 d However, in the present study we have analyzed feasibility of different spent nuclear fuel transmutation options in the fusion–fission hybrid system with conventional LiF-BeF2-(HN)F4: 28.57% – 6Li + 7Li, 14.29% – Be, 57.13% – F (F salt), which is also usually employed in ADS systems analysis (Henderson, 2011) All calculations of the flibe-based actinide transmutation blanket were made employing the Monteburns code system (Trellue and Poston, 1999) MCNP (Briesmeister, 2000) was used to obtain neutron multiplication factors keff and kscr (comprising the external source neutron input to the total neutron multiplication factor of the system) of the TRU blanket and also to estimate the neutron flux.kscr is defined as: fabrication, handling, durability, shuffling, transport, reprocessing, or fuel refabrication and radioactive waste management costs), and the lesser environmental impact (mine tailings, etc.) (Siemer, 2015) One of the reasons, among others, why GEN IV reactors are considered is the fact that they are expected to “close” the fuel cycle by significantly reducing high level waste The same applies to Accelerator Driven Systems (ADS) and fusion–fission hybrid systems, which are considered by many research organizations ADS for radioactive waste transmutation and energy production for the first time were proposed by Rubbia (1997)) and Bowman (1992) For implementation of ADS, the spallation target (neutron production) and proton beam optimization are needed together with the development of new spent fuel partitioning technique and materials technology (Henderson, 2011) The plasma-based fusion device could provide as intense neutron source as a high power accelerator up to 1019 n/s (Cheng, 1998), and this does not require very high fusion based reactors (∼200–300 MW thermal would be sufficient for our purposes) One of the attractive criteria and advantages of the fusion–fission hybrid system is that, the same as ADS, it is designed to be always sub-critical It is still too early to talk about the self-sustained fusion–fission hybrid system feasibility from the industrial point of view, but we believe that as soon as the fusion International Thermonuclear Experimental Reactor (ITER) project is completed, the demonstration of a molten-salt blanket as a medium for trans-uranium actinides (TRU) for nuclear waste transmutation could be possible as one of the applications of the fusion system – the next step for closed the fuel cycle accomplishment The progress in the plasma-based fusion device technologies is tangible: the plasma fusion driver can be designed based on latest progress in the construction and operation of the (ITER) (Merola et al., 2014) and EAST (Experimental Advanced Superconducting Tokamak) (Wu et al., 2011) There is a growing interest within the fusion community in US revisiting the concept of the fusion–fission hybrid reactor (Freidberg and Kadak, 2009; Kotschenreuthera et al., 2009) Ambitious plans have been set up to generate more than 200 GWe of nuclear power in China in 2050 and a fusion-driven subcritical system concept based on viable technologies has been proposed (Wu et al., 2009, 2011) This system is expected to recycle nuclear waste making the energy production more environmentally friendly The main parameters defining the transmutation process are the neutron energy spectrum and neutron fluxes of the system In principle, any intensive neutron source can be used for waste transmutation (Slessarev and Bolov, 2003) The more intensive neutron flux determines a shorter lifetime of a certain nucleus in the system flux On the other hand, transmutation depends on the neutron capture and the neutron-induced fission cross sections A unique solution is impossible because the capture and fission cross sections vary considerably from one isotope to another Looking for a new efficient and economically viable transmutation system the main consideration is given to the neutron flux and spectrum characteristics of a particular system (Salvatores, 2002) as well as the interaction of this neutron flux with the given composition of transmutable material For the fusion–fission hybrid system both the materials and technology development is still needed However, the neutronic characteristics in the transmutation blanket have been studied in (Cheng and Wong, 2000; Cheng, 2001; Ridikas et al., 2006) An inertial confinement fusion (ICF) device (based on the D + T → 4He+n nuclear reaction) could provide a powerful neutron source The MW fusion power corresponds to ∼4 × 1017 n/s A molten-salt blanket (LiF-BeF2(HN)F4 – “flibe”), surrounding this neutron source, then could serve as a medium for trans-uranium actinides (TRU) to be burned Flibe also has a function of both the coolant and the carrier of tritium breeding material (6Li in this case) A well known advantage of the molten salt is its possibility of both refueling of burned TRU and extraction of fission products (FP) on-line The averaged neutron flux is very high (of the order of ∼1.5·1015 n s−1 m−2) and it corresponds to the flux typical only of high-flux reactors In order to produce 3000 MWth fission kscr = (Mn−1)/(Mn−1/ ν ) (1) where ν – is the average number of neutrons per fission and Mn is a total neutron multiplication factor of the system In the scope of the EURATOM project BRILLIANT, which has been initiated to identify the barriers for developing nuclear power in the Baltic region countries and to prepare the basis for overcoming them, the alternative of closed fuel cycle technologies instead of the open cycle with the spent nuclear fuel (SNF) disposal has been analyzed One of the project objectives is to identify the milestones for the development of the sustainable nuclear fuel cycle using the transmutation technology The main aim of the work is considerations of the feasibility study of waste transmutation in the epithermal-fast neutron flux of this hybrid system by comparing transmutation scenarios of actinides extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT This research was done from the point of view of the reactor core physics and of the isotopic composition of the fuels which would ensure at least the theoretical part of the transmutation feasibility In particular, the transmutation scenario of TRU elements has been simulated by using the model of the fusion–fission hybrid system for different spent nuclear fuels accumulated in the Baltic countries: RBMK-1500 (Lithuania), BWR and PWR (Sweden) reactors In the framework of the EURATOM BRILLIANT project the results of this study provide valuable data for analysis and optimization of nuclear fuel cycle options in the Baltic region in case of the further development of nuclear power In order to investigate the transmutation efficiency of the fusion–fission hybrid system for different spent nuclear fuel TRU, the actinides extracted from the light water reactor (LWR) (averaged composition of SNF TRU) and RBMK spent fuel (different cases) were analyzed Major performance parameters of the transmutation system (e.g., kscr, keff, Φn(E), equilibrium conditions, etc.) for different TRU cases have been investigated Detailed burn-up calculations with continuous feeding of TRU material to refresh the burned one and partial treatment of fission products have been performed in the modeling 242 Nuclear Engineering and Design 330 (2018) 241–249 R Plukienė et al Table Existing PWR and BWR type nuclear reactors in Sweden and RBMK-type reactors in Lithuania The average parameters of LWR and RBMK reactors are marked in bold Name Type Fuel enrich-ment Power (MW) (PRIS) Loading factor (%) Operation time, y (until 2015 y.) Partial input to SNF Burn-up MWd/kg (U.S DOE/EIS, 2008) Forsmark Forsmark Forsmark Oskarshamn Oskarshamn Oskarshamn Ringhals Ringhals Ringhals Ringhals Barsebäck Barsebäck Sweden Ignalina-1 Ignalina-2 Lithuania BWR BWR BWR BWR BWR BWR BWR PWR PWR PWR BWR BWR LWR RBMK RBMK RBMK 2.5 2.1 2.8 2.3 2.5 2.6 2.8 3.2 2.8 4.0 2.9 3.2 2.1–4 2–2.6 2–2.8 2–2.8 984 1120 1167 473 638 1400 881 807 1063 1118 600 600 1000 1500 1500 1500 82.2 80.2 84.8 60.4 73.3 77.4 67.2 67.4 76.1 80.1 74.5 74.9 76 53.9 64.4 59.3 31.91 30.9 27.4 28.33 31.98 26.01 30.75 30.75 29.72 21.82 19.93 22.96 28 21 22 21.5 0.094 0.104 0.096 0.046 0.071 0.114 0.098 0.09 0.101 0.1 0.039 0.046 0.488 0.512 33.75 34.5 34.25 33.75 29 34.25 29 33.5 39.5 39.5 26.25 29 34 18* 20* 19 * (Shevaldin et al., 1998; Krivoshein, 2002) Initial conditions of the hybrid system and tru composition in the blanket Table Initial TRU compositions (%) in the molten salt blanket for different SNF cases The main TRU elements are marked in bold For the investigation of the possibilities to incinerate the actinides separated from SNF of different power plants in the fusion–fission hybrid system we performed an evaluation of an inventory of the spent nuclear fuel generated in Sweden and Lithuania nuclear reactors (PWR, BWR and RBMK) up to 2015 Collected data on parameters of existing PWR, BWR and RBMK-type nuclear reactors in Lithuania and Sweden are presented in Table Averaged data of the discharged fuel burn-up depending on the fuel enrichment and the discharge year in BWR and PWR reactors are taken from (U.S DOE/EIS, 2008); power and the loading factor of the reactor are taken from (PRIS website (2016)) We have used the IAEA nuclear fuel cycle simulation system (NFCSS (2012)) and a generalized reactor model assumption for the country: those different types of reactors of the same power and the same initial fuel flow will produce approximately the same amount of SNF That means that the assumption of generalized reactor model for Sweden takes into account the averaged power, the average operation time, the averaged fuel enrichment and the averaged fuel burn-up of all existing BWR and PWR nuclear reactors in the country as listed in Table According to (Plan 2013, 2014) there was 7520 tons of SNF in Sweden in 2013 According to this model the total amount of heavy metals of SNF is approximately 7630 tons including 77.5 tons of TRU at the end of 2015 During operation of the Ignalina NPP (Lithuania) about 22,000 assemblies of SNF (UO2 fuel with 2.0, 2.4, 2.6 and 2.8% initial 235U enrichment (Barkauskaset al., 2016; Shevaldin et al., 1998; Krivoshein, 2002) from the two RBMK-1500 reactors were accumulated The approximate amount of TRU is 24 tons The TRU composition in the LWR SNF significantly depends on the initial fuel enrichment and the fuel burn-up in the reactor; these parameters have changed a lot from 1974 to 2015 There exists a wide variety of fuel assembly designs used in BWR (8 × 8, × 9, SVEA-64, SVEA-100) and in PWR (15 × 15, 17 × 17) with different fuel enrichments (2.1%–4% of 235U), which also influences the TRU composition (Favalli et al., 2016) We have used the average spent nuclear fuel composition of the LWR reactor (3% enrichment 235U) after the years cooling time The isotopic composition of the average LWR reactor SNF composition was calculated by using ORIGEN-ARP (SCALE module) with specially prepared CE 14 × 14 (Westinghouse) ORIGEN-S multiburn-up libraries, which have been validated (Bowman et al., 2011, Bowman and Leal, 2000) The considered TRU isotopic compositions in the molten salt are presented in Table The RBMK-1500 SNF composition strongly depends on the initial RBMK reactor fuel enrichment and the burn-up rate (Kimtys et al., Isotope LWR RBMK RBMK RBMK 237 3.78 1.81 51.64 24.61 8.22 4.61 90.89 4.21 0.83 5.04 0.19 2.07 0.19 61.18 27.56 0.64 1.92 91.49 6.32 0.12 6.44 0.005 3.32 0.51 49.62 31.09 0.94 4.70 86.86 9.30 0.51 9.81 0.01 3.82 0.67 46.08 31.86 0.98 6.06 85.64 9.74 0.78 10.52 0.02 Np,% Pu,% 239 Pu,% 240 Pu,% 241 Pu,% 242 Pu,% Total Pu,% 241 Am,% 243 Am,% Total Am,% 244 Cm,% 238 2001, Plukienė et al., 2009) Three representative RBMK-1500 SNF of the Ignalina NPP RBMK TRU vectors after the 50 years cooling time have been evaluated: “RBMK 1” – 2% 235U enrichment and 14 MWd/kg burn-up; “RBMK 2” – 2.4% 235U enrichment, 22 MWd/kg; “RBMK 3” – 2.6% 235U enrichment, 26 MWd/kg These RBMK TRU cases have been chosen taking into account the SNF composition of the Ignalina NPP (Lithuania) The different RBMK SNF cases have been modelled using ORIGEN-ARP with prepared one-group burn-up dependent cross-section libraries, which are comprehensively explained in publication (Barkauskas et al., 2017) Depletion calculations for generation of the cross-section libraries were performed using the SCALE 6.1 code package with the TRITON control module, which employs a NEWT deterministic 2D transport code with the 238-group energy library based on ENDF-B VII library and the ORIGEN-S nuclide composition calculation code The calculated composition of actinides has been validated by comparing the evaluation against experimental data The similar isotopic composition has been obtained in benchmark calculations of the RBMK spent nuclear fuel isotopic composition using MCNP and ORIGEN codes where modeling results of 2% enrichment 235U fuel were compared with the same experimentally measured data of the RBMK-1000 fuel isotopic composition (Burlakov et al., 2003) One should note that if the RBMK-1500 SNF cooling time is 50 years the larger part of 241Pu decays to 241Am And, if the cooling time is shorter, the 241Pu amount in RBMK-1500 SNF would be closer to the LWR 241Pu amount Historically, uranium fuel with the 2% enrichment was used from the very beginning of the Ignalina NPP operation It is the same type of fuel loaded in RBMK-1000 reactors in Chernobyl (Ukraine) We have chosen two SNF burn-up cases with this enrichment: 1) 14 MWd/ kg – assemblies of low burn-up and 2) 20 MWd/kg – assemblies of high 243 Nuclear Engineering and Design 330 (2018) 241–249 R Plukienė et al continuous TRU and 6Li feeding as well as continuous FP removal have been used in all simulated cases during the irradiation A mechanism to remove the fission products is needed in the molten salt transmutation system not only due to the better neutron balance of the system – if not removed solidification of fission products will occur in the molten salt quite soon after the operation The hybrid system transmutation blanket parameters, kscr, Φn(E), equilibrium conditions and the tritium breeding ratio, have been calculated for each analyzed SNF TRU case Further, the comparative analysis of the LWR and RBMK SNF TRU incineration rate has been performed Modeling results of SNF TRU transmutation and main performance parameters of the hybrid system Here we present the results of the modeling of LWR and RBMK SNF TRU transmutation process by the previously described hybrid fusion–fission system From Table one can notice that 240Pu part is considerably larger in all RBMK TRU composition cases compared with the LWR SNF case (especially for RBMK and RBMK cases), besides the concentration of 241Am is higher and that of 244Cm is lower This is due to the lower initial RBMK fuel enrichment and higher Pu transmutation in the RBMK type reactor The strong neutron absorption by 240 Pu and a relatively smaller amount of the fissile element concentration determine the smaller kscr values in RBMK TRU as compared with the LWR TRU case (Fig 2(a)) Therefore, a larger fusion power is needed at the beginning of irradiation as it is presented in Fig 2(b) to sustain the GW thermal power of the system The fusion power is proportional to kscr: Pfus ∼ (1-kscr)/kscr By comparing LWR and RBMK TRU transmutation cases in terms of the main performance parameters of the hybrid system, a number of important advantages have been obtained in favor of the LWR isotopic composition: the values of kscr (0.83) and Pfus (180 MW) are almost stable comparing the beginning and equilibrium stages, the equilibrium is reached after ∼3 years of the system operation, and the tritium breeding rate (TBR) is sufficient to supply tritium for the fusion device (TBR = 1.25) In the case of RBMK 1, the equilibrium is reached after ∼3.3 years, kscr (0.77) is lesser at the beginning but later it approaches that of the LWR TRU case In equilibrium kscr = 0.814, Pfus = 190 MW, TBR = 1.25 In case of RBMK and RBMK transmutation in the molten salt kscr at the beginning of irradiation is 0.69 and 0.67, respectively, and the 370 MW and 440 MW fusion power is needed for corresponding cases, and the TBR is not sufficient for such fusion power (TBR < 1, 0.73 and 0.67, respectively) When the equilibrium for main isotope concentrations is reached, in both cases the situation is Fig A simplified geometry model of the fusion–fission hybrid FD stands for a Fusion Device burn-up, both representatives of a typical burn-up fuel before 1996 From 1996 up to 2004 the 2.4% 235U enrichment fuel with 0.41% burnable poison (erbium) was the most frequently used nuclear fuel in RBMK-1500; here the average fuel burn-up was 22 MWd/kg The last RBMK case was selected to test the SNF originated from later operational campaigns (from 2002) with the 2.6% 235U enrichment and 0.5% of burnable poison – erbium, which was first tested in the Ignalina NPP in ∼2001 This initial fuel load characteristic was used until the final shutdown of the Ignalina NPP (Unit I in 2004 and Unit II in 2009) The simplified spherical geometry setup described in Ridikas et al., 2006 was used in our calculations (see Fig for details) The diameter of the cavity with the fusion device is 400 cm, surrounded by the cm thick liquid flibe (6Li 0.1% in Li) wall, the 0.3 cm – metallic wall (corrosion resistant SS316 50%), and the cm thick graphite A 60 cm thick transmutation blanket (divided into regions for calculation) with TRU inside a flibe is placed between the graphite and the metallic wall All the structure was surrounded by a 20 cm thick graphite reflector and a cm thick stainless-steel shell (Cheng, 2001; Ridikas et al., 2006) Initial conditions of the hybrid system were the same for all TRU isotopic vectors considered: Ptherm GW, initial TRU mass – 3.04 t The Fig (a) kscr as a function of irradiation time in the molten salt blanket for different TRU cases, (kscr calculated with