System Studies on the Fusion-Fission Hybrid Systems and Its Fuel Cycle

applied sciences

Article System Studies on the Fusion-Fission Hybrid Systems and Its Fuel Cycle

Mikhail Shlenskii 1,* and Boris Kuteev 2,*

1 Plasma Physics Department, National Research Nuclear University Moscow Engineering Physics Institute, Kashirskoe Hwy 31, 115409 Moscow, Russia 2 Tokamak Department, National Research Center “Kurchatov Institute”, Akademika Kurchatova pl. 1, 123098 Moscow, Russia * Correspondence: [email protected] (M.S.); [email protected] (B.K.)

Received: 8 November 2020; Accepted: 24 November 2020; Published: 26 November 2020

Abstract: This paper is devoted to applications of fusion-ﬁssion hybrid systems (FFHS) as a powerful neutron source implementing transmutation of minor actinides (MA: Np, Am, Cm) extracted from the spent nuclear fuel (SNF) of nuclear reactors. Calculations which simulated nuclide kinetics for the metallic fuel containing MA and neutron transport were performed for particular facilities. Three FFHS with fusion power equal to 40 MW are considered in this study: demo, pilot-industrial and industrial reactors. In addition, needs for a ﬂeet of such reactors are assessed as well as future FFHSs’ impact on Russian Nuclear Power System. A system analysis of nuclear energy development in Russia was also performed with the participation of the FFHSs, with the help of the model created at AO “Proryv”. The quantity of MA that would be produced and transmuted in this scenario is estimated. This research shows that by the means of only one hybrid facility it is possible to reduce by 2130 the mass of MA in the Russian power system by about 28%. In the case of the absence of partitioning and transmutation of MA from SNF, 287 t of MA will accumulate in the Russian power system by 2130.

Keywords: fusion-ﬁssion hybrid system (FFHS); fusion neutron source (FNS); minor actinides; partitioning and transmutation (P&T); closed nuclear fuel cycle

1. Introduction

1.1. Prerequisites for Minor Actinides Transmutation One of the most crucial issues for nuclear engineering is to create a closed nuclear fuel cycle that would provide the use of nuclear energy for a long time (more than 1000 years) and improve the safety, ecology and economics of nuclear technology. This issue includes the problem of spent nuclear fuel (SNF) and radioactive waste (RW) management. Minor actinides (MA—Np, Am, Cm) comprise about 0.1–0.15% of SNF mass, have high radiotoxicity and a relatively long half-life [1,2]. One of the ways to manage MA is to dispose of them in geologic repositories. However, there is another approach where MA are transmuted into lighter nuclides. This paper describes the important problems of MA management and a way to solve those using FFHSs. The value of radiotoxicity shows the real harmful potential of a particular nuclide—thus it is used by a majority of authors. Radiotoxicity is a result of the multiplication of nuclide radioactivity and eﬀective dose coeﬃcient (or DPUI—Dose Per Unit Intake, Sv/Bq) [2]:

Radiotoxicity = A e(50) (1) ·

Appl. Sci. 2020, 10, 8417; doi:10.3390/app10238417 www.mdpi.com/journal/applsci Appl. Sci. 2020, 10, 8417 2 of 15 Appl. Sci. 2020, 10, x FOR PEER REVIEW 2 of 16 wherewhere A isis the nuclide’s activity activity (Bq) (Bq) for for a a particular particular moment, moment, 𝑒(50)e(50) is is thethe eeffectiveffective dosedose coecoefficientfficient (Sv(Sv/Bq)/Bq) thatthat isis thethe dosedose (Sv)(Sv) resultingresulting fromfrom thethe intakeintake ofof 11 BqBq fromfrom aa specificspecific nuclidenuclide over over 50 50 years. years. As reported in [[1,2]1,2] the typical SNF of a light wa waterter reactor (LWR) consists of more than 95% U, about 1% Pu, only 0.1% MA and about 3–4% fissionfission products (FP). Plutonium and MA (despite their small percentage)percentage) provideprovide the the main main contribution contribution to to the the long-term long-term radiotoxicity radiotoxicity of radioactiveof radioactive waste waste [2], (Figure[2], (Figure1). The 1). radiotoxicity The radiotoxicity of fission of fission products produc declinests declines much faster much compared faster compared to that of to actinides. that of Itactinides. reaches theIt reaches radioactive the radioactive equilibrium equilibrium with respect wi toth the respect uranium to the ore uranium in about 300ore yearsin about [2]. 300 If SNF years is not[2]. reprocessedIf SNF is not and reprocessed transmuted, and it reachestransmuted, the natural it reaches radiotoxicity the natural level radiotoxicity only after 100,000 level only years after [2]. At100,000 the same years time, [2]. At in the same European time, Union in the alone,European nuclear Union power alone, plants nuclear produce power annually plants produce around 2500annually tons around of SNF [25001], almost tons of the SNF same [1], amount almost the is produced same amount in the is USA produced [3]. in the USA [3].

Figure 1. Radiotoxicity ofof SNFSNF (spent(spent nuclearnuclear fuel)fuel) and and its its components components (for (for PWR, PWR, UOX, UOX, 60 60 GW GW·d/t)d/t) [1 ]. · [1]. The concept of radiation equivalent treatment of radioactive nuclides in nuclear fuel cycle [4] is quiteThe attractive concept because of radiation it could equivalent provide a treatment great reduction of radioactive of safety nuclides requirements in nuclear and costs fuel forcycle the [4] RW is repositoryquite attractive ([5], p. because 516, [6]). it Thiscould concept provide implies a great disposing reduction of of radioactive safety requirements waste that has and a radiotoxicitycosts for the levelRW repository equal to the ([5], natural p. 516, U or [6]). Th oreThis level concept (or at implies least reduced disposing radiotoxicity of radioactive level). waste that has a radiotoxicityHowever, level nowadays, equal to manythe natural EU countries U or Th (andore level not just(or at EU least countries) reduced have radiotoxicity chosen a level). phase-out scenarioHowever, for nuclear nowadays, energy many (like Sweden EU countries [7] and (and Germany) not just and EU/or countries) have decided have not chosen to reprocess a phase-out SNF andscenario simply for tonuclear dispose energy of it (like in geologic Sweden repositories [7] and Germany) (like the and/or USA have [1,7]). decided There arenot severeto reprocess problems SNF withand simply this approach. to dispose It requiresof it in geologic controlling repositories SNF for an (like impractical the USA length[1,7]). There of time, are meaning severe problems it is very expensivewith this approach. (due to the It requires time and controlling storage space SNF needed), for an impractical probably unsafe,length of not time, to mentionmeaning the it is direct very disposalexpensive of (due SNF willto the deprive time and humanity storage of space a great needed), source ofprobably power. unsafe, not to mention the direct disposalBecause of SNF of nuclearwill deprive power humani RW problem,ty of a great “pure” source thermonuclear of power. fusion reactors seem like a good alternativeBecause as of they nuclear produce power much RW less problem, RW during “pure” their th operationermonuclear if the fusion correct reactors materials seem are like chosen a good [8]. However,alternative the as fusionthey produce reactor ismuch a very less complex RW during facility their that operation is in fact ine if ffithecient correct in utilizing materials very are valuable chosen thermonuclear[8]. However, the neutrons. fusion Atreactor the same is a timevery fusion-fission complex facility hybrid that systems is in fact have inefficient relatively in easily utilizing achieved very plasmavaluable parameters thermonuclear compared neutrons. to “pure” At the thermonuclear same time fusion-fission fusion reactors hybrid [9]. systems have relatively easilyThus achieved reprocessing plasma parameters the spent nuclear compared fuel to via “pure” separating thermonuclear minor actinides, fusion reactors Pu and [9]. U seems a promisingThus reprocessing way forward the in developmentspent nuclear of fuel the via nuclear separating power. minor Pu and actini U coulddes, Pu be and used U in seems fission a reactorspromising again. way forward Fission productsin development could beof usedthe nuclear as sources power. of radiation,Pu and U reducingcould be theused amount in fission of radioactivereactors again. waste. Fission As mentioned products above, could FPbe lose used their as radiotoxicitysources of radiation, much faster reducing than actinides. the amount The last of radioactive waste. As mentioned above, FP lose their radiotoxicity much faster than actinides. The last problem, concerning MA, could be solved via the transmutation of MA in fission reactions which produce fission products. This strategy is called partitioning & transmutation (P&T) [1]. Appl. Sci. 2020, 10, 8417 3 of 15 problem, concerning MA, could be solved via the transmutation of MA in fission reactions which Appl. Sci. 2020, 10, x FOR PEER REVIEW 3 of 16 produce fission products. This strategy is called partitioning & transmutation (P&T) [1]. AA very very hard hard spectrum spectrum is requiredis required for for the the fission fission of most of most MA dueMA todue the to threshold the threshold character character of their of fissiontheir fission cross-sections cross-sections (Figure 2(Figure). In Figure 2). 2In, twoFigure types 2, oftwo neutron types cross-sectionof neutron cross-section are shown: fission are shown: and capture.fission and The capture. ratio of theirThe ratio average of their values average for a particularvalues for neutrona particular spectrum neutron (α )spectrum is crucial ( toα) eisff ectivecrucial transmutation.to effective transmutation. Table1 shows Table that the 1 shows capture-to-fission that the capture-to-fission ratio for the spectrum ratio for of the a light spectrum water of reactor a light iswater bigger reactor than 1is forbigger most than actinides 1 for most while actinides for the fastwhile reactor for the spectrum fast reactor this spectrum ratio less this than ratio 1 for less most than actinides.1 for most At actinides. the same At time, the αsameis much time, greater α is much than greater 1 for both than of 1 the for spectrums both of the in spectrums the cases ofin Am-241,the cases Np-237of Am-241, and Am-243,Np-237 and which Am-243, comprise which the comprise main part the of MA.main part of MA.

FigureFigure 2. 2.Microscopic Microscopic cross-sections cross-sections of of Np-237 Np-237 and and Am-241 Am-241 against against incident incident neutron neutron energy. energy.

Table 1. Average ﬁssion, capture and capture-to-ﬁssion ratios α for selected TRU isotopes *. Table 1. Average fission, capture and capture-to-fission ratios α for selected TRU isotopes *. R R σ = σ(E)ϕ(E)dE/ ϕ(E)dE 𝝈 𝝈(𝑬)𝝋(𝑬)𝒅𝑬 / 𝝋(𝑬)𝒅𝑬 For PWR Spectrum For Fast Reactor Spectrum Isotope σFor PWRσ Spectrumα For σFast Reactorσ Spectrumα Isotope f, b c, b f, b c, b σf, b σc, b α σf, b σc, b α U-235 39.86 9.25 0.23 1.95 0.56 0.29 U-238U-235 39.86 0.10 9.254.46 0.23 42.87 1.95 0.05 0.56 0.34 0.29 7.43 Np-237U-238 0.530.10 4.4635.37 42.87 66.84 0.05 0.34 0.34 1.80 7.43 5.24 Np-238Np-237 120.500.53 35.3731.03 66.84 0.26 0.34 1.71 1.80 0.20 5.24 0.12 Pu-238Np-238 120.50 2.32 31.0320.35 0.26 8.78 1.71 1.20 0.20 1.04 0.12 0.86 Pu-239 105.30 60.07 0.57 1.83 0.54 0.30 Pu-240Pu-238 0.632.32 20.35233.30 8.78 372.92 1.20 0.40 1.04 0.57 0.86 1.42 Pu-241Pu-239 105.30 105.00 60.0738.47 0.57 0.37 1.83 2.54 0.54 0.48 0.30 0.19 Pu-242Pu-240 0.440.63 233.3029.96 372.92 68.18 0.40 0.28 0.57 0.40 1.42 1.45 Am-241Pu-241 105.00 1.11 38.47111.01 0.37 100.37 2.54 0.28 0.48 2.00 0.19 7.26 Am-242 195.20 Pu-242 0.44 29.9628.37 68.18 0.15 0.28 3.21 0.40 0.39 1.45 0.12 Am-242m 469.00 105.40 0.22 3.31 0.21 0.06 Am-243Am-241 0.471.11 111.0151.65 100.37 110.25 0.28 0.21 2.00 1.56 7.26 7.54 Cm-242Am-242 195.20 1.37 28.374.02 0.15 2.93 3.21 0.63 0.39 0.49 0.12 0.77 Cm-243Am-242m 469.00 79.07 105.4012.10 0.22 0.15 3.31 2.82 0.21 0.28 0.06 0.10 Cm-244Am-243 0.900.47 51.6517.77 110.25 19.85 0.21 0.49 1.56 0.64 7.54 1.31 Cm-245 119.80 19.22 0.16 2.72 0.41 0.15 Cm-242 1.37 4.02 2.93 0.63 0.49 0.77 * The values wereCm-243 obtained 79.07 using TENDL_2014,12.10 0.15 FISPACT-II 2.82 and reference 0.28 incident 0.10 particle spectra (PWR-UO2-0, Superphenix). Cm-244 0.90 17.77 19.85 0.49 0.64 1.31 Cm-245 119.80 19.22 0.16 2.72 0.41 0.15 * The values were obtained using TENDL_2014, FISPACT-II and reference incident particle spectra (PWR-UO2- 0, Superphenix) Appl. Sci. 2020, 10, 8417 4 of 15 Appl. Sci. 2020, 10, x FOR PEER REVIEW 4 of 16

AnotherAnother issue issue connected connected with with MA MA transmutation transmutation is is the the low low fraction fraction of of delayed delayed neutrons neutrons for for MA MA nuclidesnuclides (Table (Table2)[ 2)1 [1].]. It It is is an an obstacle obstacle to to MA MA transmutation transmutation in in a criticala critical reactor. reactor.

TableTable 2. 2.The The e effectiveﬀective fraction fraction of of delayed delayed neutrons. neutrons.

NuclideNuclide ββ 238238UU 0.01720.0172 237237NpNp 0.003880.00388 238238PuPu 0.001370.00137 239239PuPu 0.002140.00214 240240PuPu 0.003040.00304 241241PuPu 0.005350.00535 242 242PuPu 0.006640.00664 241 241AmAm 0.001270.00127 243 243AmAm 0.002330.00233 242 242CmCm 0.0003770.000377

OneOne more more feature feature of allof actinidesall actinides that isthat important is important for neutronics for neutronics is the number is the ofnumber prompt of neutrons prompt perneutrons ﬁssion per that fission increases that withincreases initial with neutron initial energyneutron growth energy (Figure growth3 ).(Figure Thus the3). Thus higher the the higher initial the neutroninitial neutron energy, energy, the more the eﬀ moreective effective the neutron the neutron utilization. utilization.

FigureFigure 3. 3.Number Number ofof promptprompt neutronsneutrons perper ﬁssion fission against against initial initial neutron neutron energy. energy.

TheThe neutronics neutronics features features of of MA MA discussed discussed above above are are the the reasons reasons why why subcritical subcritical systems systems are are more more appropriateappropriate for for MA MA transmutation transmutation compared compared to critical to critical reactors. reactors. One of the One most of promising the most approaches promising toapproaches solving this to problem solving isthis the problem application is the of a application fusion reactor of asa fusion a high reactor energy neutronas a high source. energy This neutron idea issource. investigated This idea in [ 1is,3 investigated,10–12]. in [1,3,10–12].

1.2. Studies on FFHS as Burner Reactor The majority of foreign researchers consider Pu to be a problem like MA and even more serious because of the proliferation risks. The author of [3] is not an exception: he investigates the transmutation of fuel that consists of MA and Pu in diﬀerent FFHSs (with blanket thermal power from 3 to 10 GW). Appl. Sci. 2020, 10, 8417 5 of 15

FFHS do not have high requirements for plasma parameters because most of the neutrons are born inside a blanket (that contains fissile materials). For example, if keff = 0.95 then the neutron multiplication factor for a plasma source will be about 20. The fusion power of these FFHSs varies between 150 MW and 500 MW (that corresponds to ITER fusion power). The author of [3] pointed out that for FFHSs the following parameters are appropriate: fusion power <200 MW, normalized beta < 2, neutron wall 2 loading < 1 MW/m , power amplification Qp < 2. In this case, the corresponding efficiency of burning Pu-MA fuel is 1.1 A, 1.65 A, 3.7 A tons annually (where A is availability factor). The original × × × values were changed according to Pu + MA fraction in the SNF (1.1%). The authors of [10] point out the advantages of a FFHS for MA transmutation and investigate the utilization of a thermonuclear neutron source with fusion power less than 200 MW for the transmutation of the metal fuel consisting of MA, Pu and Zr. The proportions of fuel components are varied: the more Pu there is in the fuel the lower MA transmutation rate, but the higher blanket thermal power. It was found that the transmutation rate of MA could reach 19.4% for each irradiation cycle (that is 5 y) and about 86.5% for 25 cycles. Another approach for MA transmutation in a FFHS is an application of a molten salt blanket where actinide fluorides are dissolved in a molten eutectic of alkali metal fluorides. This salt also serves as a coolant. It is a very promising blanket option, having, for example, such advantages as online fuel composition control. Such an approach is addressed in [11]. In this study, the authors decided to take all transuranic elements (TRU) from SNF (i.e., Pu and MA) for fuel fabrication. The fusion power of this reactor is 250 MW and the blanket thermal power is 3 GW. In this case, it’s possible to burn about 1.1 t of TRU annually. The authors also stressed that removing more than 50% of FP from the fuel after each irradiation cycle is vital. Finally, not only are efficiency parameters important, but also the impact of FFHSs on a nuclear fuel cycle. This impact was successfully established by research [12] for Japan’s nuclear power. This paper addresses the utilization of three stellarators that have extremely high fusion powers (as for FFHS) of 1, 2 and 3 GWt. These reactors are assumed to be introduced in 2050, 2070 and 2090 respectively. The authors demonstrate that a FFHS with 1 GWt fusion power is able to burn about 350 kg of MA per year, which seems like a very small amount compared to other studies. There is a discrepancy between [11] and [12] on the impact of FPs’ accumulation on the MA burning process. Thus a more detailed analysis of this question is required for new studies. The authors of [12] also show that all MA in the system could be burned by utilization of the stellarators mentioned above in approximately 200 years.

1.3. Conclusions on Literature Review and Details of This Study As can be seen in the presented studies, FFHSs are a really promising solution for the transmutation of TRU that would help to manage the problem of nuclear waste storage. Although there is a considerable number of studies on the application of FFHSs for MA transmutation, the majority of them are only at the beginning stage of designing a real facility or have no intention of designing real facilities at all. Another drawback of many studies is the lack of a comprehensive analysis of the problem: not only are the parameters of the transmutation essential, but also comparison with other competing approaches to the transmutation and the assessment of the impact on the nuclear power system. Our team has already completed many steps of the FFHS’s design. We use the ITER project as a foundation for our own. New studies are needed for particular FFHS implementation projects. At present, the demo FFHS (DEMO-FNS) is at the design stage. The road map of the project also assumes pilot industrial (PIHR) and industrial (IHR) hybrid reactors. All these reactors have identical fusion power (40 MW), similar construction, different blanket fuel loadings and capacity factors. In this paper, an assessment of the potential of the assumed hybrid reactors for MA incineration is performed. Performance parameters are compared for different coolants (H2O and CO2) in the transmutation zone. Features of fuel inventory evolution during transmutation are analyzed. Another Appl. Sci. 2020, 10, 8417 6 of 15 Appl. Sci. 2020, 10, x FOR PEER REVIEW 6 of 16 importantIn aimthis paper, of this an study assessment is to estimate of the potential the accumulated of the assumed amount hybrid of MAreactors in Russian for MA incineration nuclear power systemis performed. over the nextPerformance 100 years parameters and FFHSs’ are compared potential for different reducing coolants this amount. (H2O and For CO this2) in purpose, the the Universaltransmutation System zone. Model Features of Russian of fuel nuclear inventory power evolution was used during [13]. transmutation are analyzed. AnotherThere are important two sections aim of in this this study study. is to The estimate firstis the a studyaccumulated of fuel amount inventory of MA evolution in Russian in thenuclear blanket of FFHSspower containing system over a metallicthe next fuel100 thatyears is and a zirconium FFHSs’ potential and MA for alloy. reducing The secondthis amount. part ofFor the this study purpose, the Universal System Model of Russian nuclear power was used [13]. is a system analysis of the development of nuclear power in Russia and the involvement of the There are two sections in this study. The first is a study of fuel inventory evolution in the blanket hybrid reactors. of FFHSs containing a metallic fuel that is a zirconium and MA alloy. The second part of the study is a system analysis of the development of nuclear power in Russia and the involvement of the hybrid 2. Materials and Methods reactors. All three hybrid reactors, assumed by the road map of the NRC “Kurchatov Institute” project, will2. be Materials based on and a tokamak Methods with a blanket containing fissile materials and lithium. TheAll demo three hybrid hybrid reactorreactors, DEMO-FNSassumed by the is road the firstmap ofof the NRC reactors “Kurchatov mentioned Institute” above, project, and its contemporarywill be based design on a tokamak is the most with detaileda blanket andcontaining defined. fissile The materials other two and reactorslithium. will have a similar constructionThe withdemo slightly hybrid direactorfferent DEMO-FNS parameters. is Allthe of firs thet of tokamaks the reactors of the mentioned hybrid systems above, consideredand its havecontemporary the following design identical is the parameters: most detailed major and defined. plasma radiusThe otherR0 two= 320 reactors cm, minor will have plasma a similar radius a construction with slightly different parameters. All of the tokamaks of the hybrid systems considered = 100 cm, plasma current Ip = 5 MA, toroidal magnetic field Bt0 = 5 T, fusion power Pfus = 40 MW (thathave corresponds the following to ~1.4 identical10 parameters:19 n/s for D-T major reaction) plasma andradius eff Rective0 = 320 fuel cm, irradiationminor plasma time radius is 5 ayears. = × Each100 reactor cm, plasma is due current to commence Ip = 5 MA, operations toroidal magnetic in 2033, 2045field andBt0 = 20555 T, fusion respectively. power P Thefus = 40 di ffMWerences (thatthat corresponds to ~1.4·1019 n/s for D-T reaction) and effective fuel irradiation time is 5 years. Each reactor are important for this study are listed below in Table3: is due to commence operations in 2033, 2045 and 2055 respectively. The differences that are important for this study are listed below in Table 3: Table 3. Capacity factor and fuel loading with MA (minor actinides).

Table 3. CapacityParameter factor and fuel loading with DEMO-FNS MA (minor actinides). PIHR IHR CapacityParameter factor DEMO-FNS 0.3 PIHR 0.8 IHR 0.95 Fuel loading (MACapacity+ Zr), factor t (H2O—coolant) 26.24 0.3 0.8 26.24 0.95 41.68 Fuel loading (MA + Zr), t (H2O—coolant) 26.24 26.24 41.68

TheThe 3D geometrical3D geometrical model model of of DEMO-FNS DEMO-FNS forfor neutronneutron transport transport calculation calculation via via the the Monte-Carlo Monte-Carlo methodmethod is shownis shown in Figuresin Figures4 and 4 and5. The5. The blanket blanket contains contains 1818 assembliesassemblies with minor minor actinides actinides (transmutation(transmutation area), area), with with lithium lithium salt salt in in the the remaining remaining spacespace (tritium (tritium breeding breeding area). area).

Figure 4. Horizontal equatorial cut of DEMO-FNS (fusion neutron source) model. TFC—Toroidal ﬁeld coil. Appl. Sci. 2020, 10, x FOR PEER REVIEW 7 of 16

Appl. Sci. 2020, 10, 8417 7 of 15 Figure 4. Horizontal equatorial cut of DEMO-FNS (fusion neutron source) model. TFC—Toroidal field coil.

Figure 5. Horizontal equatorial cut of DEMO-FNS model—transmutation area. Figure 5. Horizontal equatorial cut of DEMO-FNS model—transmutation area.

Two coolants for the transmutation area are examined: CO2 and H2O. The coolant flows Two coolants for the transmutation area are examined: CO2 and H2O. The coolant flows vertically, vertically, along the fuel rods, inside the assemblies’ casings. H2O is chosen as a basic option for the along the fuel rods, inside the assemblies’ casings. H O is chosen as a basic option for the blanket. 2 3 blanket. The average coolant density inside the assemblies for H2O is 0.373 g/cm and for CO2 is 0.14 3 The average coolant density inside the assemblies for H2O is 0.37 g/cm and for CO2 is 0.14 g/cm . g/cm3. Utilization of CO2, instead of H2O, causes keff growth (up to 1.04). It is thus necessary to Utilization of CO2, instead of H2O, causes keff growth (up to 1.04). It is thus necessary to decrease total decrease total fuel loading in the case of CO2 as a coolant to 19.7 t (for DEMO-FNS). fuel loadingA metal in thealloy case of MA of CO and2 asZr awas coolant chosen to as 19.7 a fuel. t (for This DEMO-FNS). alloy has a theoretical density of 15 g/cm3. 3 ManyA metal researchers alloy of have MA andconsidered Zr was this chosen type as of afuel fuel. [10,14,15], This alloy and has it ahas theoretical even been density utilized of in 15 a gfast/cm . Manyreactor researchers [16]. A detailed have considered fuel inventory this is type shown of fuel in Table [10,14 4., 15], and it has even been utilized in a fast reactor [16]. A detailed fuel inventory is shown in Table4. Table 4. Fuel inventory (15 g/cm3), % (mass). Table 4. Fuel inventory (15 g/cm3), % (mass). Nuclide Fraction, % Nuclide Fraction, % NuclideNp-237 Fraction,28.67 % NuclideZr-90 Fraction, 2.05 % Am-241 62.10 Zr-91 0.45 Np-237 28.67 Zr-90 2.05 Am-242m 0.06 Zr-92 0.70 Am-241 62.10 Zr-91 0.45 Am-243 4.63 Zr-94 0.72 Am-242m 0.06 Zr-92 0.70 Am-243Cm-244 4.630.49 Zr-94Zr-96 0.12 0.72 Cm-244MA 0.4995.96 Zr-96Zr 4.04 0.12 MA 95.96 Zr 4.04 Calculations on nuclide kinetics were carried out by the program FISPACT-II [17] using a constant neutron spectrum. The neutron spectrum for the transmutation area was obtained as a result ofCalculations neutron transport on nuclide calculations kinetics using were the carried Mont oute-Carlo by the method. program The FISPACT-II spectrum is [17 volume-averaged] using a constant neutronfor the spectrum. whole transmutation The neutron area. spectrum The utilization for the transmutation of a constant area neutron was obtainedspectrum asfor a calculations result of neutron on transportnuclide calculations kinetics gives using some the error Monte-Carlo in the results, method. because The of spectrum the impact is volume-averagedof fuel inventory evolution for the whole on transmutationthe spectrum area. and Thevice utilizationversa. However, of a constant in the case neutron of a spectrumsubcritical for system calculations with an onexternal nuclide neutron kinetics givessource, some this error error in theshould results, not becausebe very high. of the Diff impacterent of researchers fuel inventory address evolution contradictory on the spectrumassessments and viceon versa. this error However, [11,12]. in the case of a subcritical system with an external neutron source, this error should notThe beneutron very high. data Dilibraryfferent ENDF/B-VI researchers is addressused for contradictory neutron transport assessments calculation. on this The error spectrum [11,12 ]. wasThe obtained neutron for data 709 library energy ENDF groups/B-VI that is correspon used for neutrond to the CCFE-709 transport calculation.group structure. The spectrumThe nuclear was obtained for 709 energy groups that correspond to the CCFE-709 group structure. The nuclear data library TENDL_2014 is used for nuclide kinetics calculations. This explains the reason for utilizing the CCFE-709 group structure for the neutron spectrum. Appl. Sci. 2020, 10, x FOR PEER REVIEW 8 of 16

Appl. Sci. 2020, 10, 8417 8 of 15 data library TENDL_2014 is used for nuclide kinetics calculations. This explains the reason for utilizing the CCFE-709 group structure for the neutron spectrum. 3. Results and Discussion 3. Results and Discussion 3.1. Neutron Transport and keff Calculations 3.1. Neutron Transport and keff Calculations The neutron energy spectra in the transmutation area of DEMO-FNS with different coolants are shownThe in Figureneutron6: energy spectra in the transmutation area of DEMO-FNS with different coolants are shown in Figure 6: 14 2 1 • In the case of H2O as a coolant, the volume-averaged total neutron flux is Φn = 2.88 10 (cm s)− ,14 • In the case of H2O as a coolant, the volume-averaged total neutron flux is ×𝛷 = 2.88 ×· 10 average2 − neutron1 energy is En = 3.52 MeV and k = 0.95. (cm ·s) , average neutron energy is 𝐸 = 3.52 eMeVff and keff = 0.95. 14 2 1 • In the case of CO2,2Φ𝛷n = 3.02 10 14 (cm 2 s)−1 , 𝐸E n = 3.47 MeV and keffeff = 0.91. • In the case of CO , = 3.02× × 10 (cm ··s) , = 3.47 MeV and k = 0.91. AsAs follows follows fromfrom the obtained spec spectra,tra, 99% 99% of of neutrons neutrons have have energy energy higher higher than than 0.001 0.001 MeV, MeV, and and40% 40% of them of them have have energy energy higher higher than than 1 1MeV. MeV. 1 1Me MeVV is is the the threshold threshold fission fission energyenergy forfor mostmost ofof the the actinidesactinides that that have have a a threshold threshold type type fission fission cross-section cross-section (Figure (Figure2). 2). As As follows follows from from this this distribution, distribution, DEMO-FNSDEMO-FNS has has quite quite a a hard hard spectrum spectrum in in the the transmutation transmutation area, area, which which is is a a positive positive feature feature for for MA MA transmutation.transmutation. The The di differencesfferences between between the the two two spectra spectra reveal reveal that that in in the the case case of of CO CO2,2, the the spectrum spectrum becomesbecomes harder, harder, the the number number of of fusion, fusion, and and especially especially fission, fission, neutrons neutrons rises. rises. It It is is caused caused by by a a decrease decrease inin coolant coolant density density (from (from 0.37 0.37 g/cm g/cm3 to3 0.14to 0.14 g/cm g/cm3) and3) and moderating moderating efficiency efficiency (from (from 1.7 1.7·10104 to42.1 to 2.1·10103 3 × × forfor 14 14 MeV MeV neutrons) neutrons) and and also also by by a a decrease decrease in in the the volume volume percentage percentage of of the the fuel fuel in in the the assemblies. assemblies.

FigureFigure 6. 6.Normalized Normalized neutron neutron spectrum spectrum in in the the transmutation transmutation area area of of DEMO-FNS. DEMO-FNS.

3.2.3.2. Neutron Neutron Cross-Sections Cross-Sections Analysis Analysis TheThe obtainedobtained spectraspectra werewere usedused forfor calculating calculating thetheaverage averagecross-sections cross-sections byby meansmeans ofofthe the FISPACT-IIFISPACT-II program. program. AsAs shownshown inin FigureFigure7 ,7, there there are are three three important important actinides actinides (Np-237, (Np-237, Am-241, Am-241, Am-243)Am-243) that that have have a a capture capture to to fission fission ratio ratioα αhigher higher than than 1 1 despite despite there there being being very very hard hard neutron neutron E spectraspectra inside inside the the transmutation transmutation area area ( (𝐸n> >3 3 MeV). MeV). At At the the same same time, time, it it is is necessary necessary to to highlight highlight that that Appl. Sci. 2020, 10, 8417 9 of 15 Appl. Sci. 2020, 10, x FOR PEER REVIEW 9 of 16 thisthis ratioratio isis justjust slightly slightly higherhigher thanthan 1,1, comparedcompared toto the the same same ratio ratio for for light-water light-water andand fastfast reactors reactors (Table(Table1 ).1). This This feature feature of ofα αfor for some some of of the the actinides actinides has has the the following following explanation. explanation. MostMost MAMA havehave almostalmost equal equal fission fission cross-section cross-section which which is higher is higher than than capture capture cross-section cross-section for neutron for neutron energy higherenergy thanhigher 1 MeV, than while 1 MeV, for while some offor them some (Np-237, of them Am-241, (Np-237, Am-243), Am-241, the Am-243), capture cross-sectionthe capture cross-section is significantly is largersignificantly than the larger fission than cross-section the fission forcross-section neutron energy for neutron less than energy 1 MeV. less Atthan the 1 same MeV. time, At the 60% same of neutronstime, 60% have of neutrons energy between have energy 0.001 and between 1 MeV. 0.001 This causesand 1 anMeV. offset This of thecauses average an offset fission of cross-section the average tofission a low cross-section energy region to where a low itenergy is less region than the where capture it is cross-section.less than the capture cross-section. AsAs follows follows from from Figure Figure7, CO7, CO2 is2 theis the best best option option for MAfor MA transmutation transmutation as for as this for coolant this coolant all main all actinidesmain actinides have the have lowest the lowestα value. α However,value. However, there is there a problem is a problem for MA for transmutation MA transmutation in the case in the of thiscase coolant of this thatcoolant will that be shown will be below. shown below.

Figure 7. Average capture-to-fission ratios α for different coolants. Figure 7. Average capture-to-ﬁssion ratios α for diﬀerent coolants.

3.3.3.3. ResultsResults ofof NuclideNuclide KineticsKinetics AnalysisAnalysis ForFor allall threethree reactors,reactors, thethe basicbasic eeffectiveﬀective irradiation irradiation timetime ofof thethe fuelfuel isis 5 5 years. years. However,However, duedue toto thethe capacitycapacity factor,factor, thethe total total time time of of idle idle plus plus irradiation irradiation isis longer.longer. ForFor thethe demodemo reactor,reactor, itit isis aboutabout 16.716.7 years years (the (the capacity capacity factor factor is is 0.3). 0.3). AllAll ofof thethe mainmain parametersparameters ofof thethe transmutationtransmutation forfor eacheach reactorreactor arearelisted listedin inTable Table5 .5. This This table table showsshows inin particular particular the the mass mass changing changing in in the the most most important important actinides actinides during during irradiation. irradiation. ThereThereare are massmass ﬂuctuationsfluctuations duringduring thethe reactor’sreactor’s operatingoperating cyclecycle causedcaused byby thethe alternatingalternating shiftsshifts betweenbetween irradiationirradiation and and cooling cooling time time (idle). (idle).

3.3.1. DEMO-FNS and H2O as a Coolant For DEMO-FNS, which has light water as a coolant, the mass of all MA decreases whereas the mass of U and Pu isotopes increases. The accumulation of other actinides is less than 1 g/t. DEMO-FNS has a low eﬃciency of actinide incineration (just 54.3 kg/year) due to the low capacity factor. A more detailed analysis of the ﬁnal fuel inventory shows that the biggest fraction of Pu is made by Pu-238 (80%), Pu-242 (11%), and Pu-240 (6.5%). Almost all U is U-234 (99%). Despite Cm mass Appl. Sci. 2020, 10, 8417 10 of 15 decreasing in total, new isotopes appear: Cm-242 (12%) and Cm-245 (2%), due to neutron capture on Am and initial Cm-244. The main pathways to Pu-238 begin from Np-237 (37%) and Am-241 (60%) that undergo neutron capture. U-234 is a result of Pu-238 α-decay.

Table 5. Main parameters of MA transmutation in FFHSs.

DEMO-FNS DEMO-FNS PIHR IHR (H2O) (CO2) Total fuel loading, t 26.24 19.7 26.24 41.68 Irradiation + Idle, y 16.7 16.7 6.25 5.25 Np 9.9 (3.5%) 5.2 (1.8%) 19.1 (6.7%) 100.3 (35.0%) − − − − Change in actinides Am 70.2 (10.5%) 63.2 (9.5%) 60.8 (9.1%) 241.8 (36.2%) − − − − mass: accumulation (+); Cm 0.4 (7.7%) 1.0 (21.2%) +2.8 (57.5%) +17.7 (358.1%) reduction (–), kg/t (per − − ton of the fuel) U +2.4 +1.6 +0.9 +3.0 Pu +43.5 +29.5 +41.7 +174.0 Burnup of actinides, %(mass) 3.6 4.0 3.6 15.4 Total actinides incineration, kg 906.4 755.0 904.4 6148.0 Eﬃciency of actinides incineration, kg/year 54.3 45.2 144.7 1171.0 Time-averaged ﬁssion power, MW(th) 472 398 472 3100

3.3.2. DEMO-FNS and CO2 as a Coolant

As mentioned above, simply changing the coolant to CO2 in the transmutation area causes keff growth up to 1.04, thus reducing the total fuel loading for this option to 19.7 t. With CO2, the qualitative change in mass of all actinides is the same as with H2O, but the quantitative parameters are different. Burnup of actinides increases by 14.3%, however the total actinides incineration is less because of a lower fuel loading. The amount of transmuted Np and Am is less, as well as U and Pu accumulation, while the amount of transmuted Cm is two times larger. This change is caused by a harder spectrum in the case of CO2, so neutron capture cross-sections decrease more than fission cross-sections increase. Thus total cross-sections decrease, and the transmutation rate reduces as a result. At the same time, due to the lower neutron capture rate, the accumulation of Cm is less than before and its fission rate is very high (its isotopes have low α (Figure7)). Thus, the transmuted amount of Cm is larger than before. In the initial fuel composition, there is a small amount of Cm, so despite its isotopes experiencing fission at a high rate, new isotopes of Cm appear due to the neutron capture reaction on other actinides. The amount of Cm may even increase (as in [18]): the neutron capture rate for other actinides is higher than the fission rate of Cm. This situation is typical for the operation of both the pilot industrial (PIHR) and industrial (IHR) reactors.

3.3.3. Pilot Industrial Hybrid Reactor Compared to DEMO-FNS, PIHR has a shorter operating cycle, accounting for a higher capacity factor (0.8). Due to this circumstance, a smaller amount of Am-241 decays into Np-237, thus the transmuted amount of Np is greater while the accumulated amount of U is smaller (U-234 is a product of Np-237 decay). The eﬃciency of actinides’ incineration is higher for PIHR than for the demo reactor, but the total incinerated amount of actinides is almost the same.

3.3.4. Industrial Hybrid Reactor—The Final Aim The final purpose of the FFHS project is to create an industrial hybrid reactor. IHR will have identical constructions with DEMO-FNS and PIHR, 40 MW of fusion power, but a larger fuel loading (40 t). This causes a growth of the neutron flux in the transmutation area. For this study, it is assumed that the neutron flux will increase five times compared to the demo reactor using H2O as a Appl. Sci. 2020, 10, x FOR PEER REVIEW 11 of 16

3.3.4. Industrial Hybrid Reactor—The Final Aim The final purpose of the FFHS project is to create an industrial hybrid reactor. IHR will have identical constructions with DEMO-FNS and PIHR, 40 MW of fusion power, but a larger fuel loading Appl. Sci. 2020, 10, 8417 11 of 15 (40 t). This causes a growth of the neutron flux in the transmutation area. For this study, it is assumed that the neutron flux will increase five times compared to the demo reactor using H2O as a coolant. Thecoolant. high Thecapacity high capacityfactor (0.95) factor and (0.95) large and fuel large loading fuel make loading this make reactor this effective. reactor e ﬀNotective. only Not for onlyMA transmutation,for MA transmutation, but also but for also electrical for electrical power power production production exceeding exceeding the the reactor’s reactor’s operational consumption, so this facility will solve the problem of MA transmutation not not only from a technical, but also from an economical, perspective. The evolution of the fuel inventory during the operatingoperating cycle for IHR is similar to that of PIHR but, due to a higher neutron flux, ﬂux, th thee transmutation rate is higher. This reactor will be able to produce more than 3 GW(th) and ~1 GW(el), while its own electrical energy consumption is about 200 MW.

3.4. FFHS in Two-Component Nuclear Power System of Russia and Its Potential forfor MAMA ReductionReduction In thethe firstfirst part part of of this this study, study, the potentialthe potentia of FFHSsl of FFHSs for MA for transmutation MA transmutation was defined. was However,defined. whatHowever, also matterswhat also is the matters impact is of the FFHSs’ impact operation of FFH onSs’ theoperation nuclear poweron the systemnuclear that power reduces system the totalthat reducesamount ofthe MA. total For amount this purpose, of MA. theFor Universalthis purpose, System the ModelUniversal designed System by Model E. V.Muraviev designed [by13 ]E. was V. Muravievused. This [13] model was describes used. This the model history describes and development the history plansand development of Russian nuclear plans of power Russian in terms nuclear of productpower in flows terms (fuel, of product waste, energy,flows (fuel, money) waste, between energy, important money) components between important of the system components (power plants,of the systemfuel fabrication (power plants, and reprocessing fuel fabrication plants, and consumers). reprocessing The plants, principal consumers). scheme of The this principal model is scheme shown of in thisFigure model8. The is modelshown covers in Figure a period 8. The of model time, fromcovers 1970 a period to 2130, of andtime, has from the 1970 following to 2130, features: and has the following features: The two-component structure of nuclear power system for the 21st century, i.e., there will be two • • typesThe two-component of reactors: thermal structure spectrum of nuclear and fast power spectrum system for the 21st century, i.e., there will be Installedtwo types capacity of reactors: of nuclear thermal power spectrum plants and in 2100fast spectrum is 92 GW(e), and 107 GW(e) in 2130 • TheInstalled accumulated capacity amountof nuclear of SNFpower from plants VVER in 2100 is ~5000 is 92 t GW(e), and 107 GW(e) in 2130 • VVERThe accumulated commissioning amount in Russia of SNF will from end VVER in 2040 is ~5000 t • VVER commissioning in Russia will end in 2040 The export of nuclear power plants and their services are not taken into account • The export of nuclear power plants and their services are not taken into account

Figure 8. Fusion neutron neutron source in the structure of Russian nuclear power system (The Universal System Model).

Figure9 shows the past and future planned evolution of the net installed capacity of nuclear power plants in Russia. In 2100, the last thermal reactor will be decommissioned and the system will consist of only fast reactors. Appl. Sci. 2020, 10, x FOR PEER REVIEW 12 of 16

Figure 9 shows the past and future planned evolution of the net installed capacity of nuclear Appl.power Sci. 2020 plants, 10, 8417 in Russia. In 2100, the last thermal reactor will be decommissioned and the system 12will of 15 consist of only fast reactors.

FigureFigure 9. 9.Evolution Evolution of of the the net net installedinstalled capacity of nuclear nuclear power power plants plants in in Russia. Russia.

InIn this this part part of of the the investigation, investigation, some some parametersparameters of the the FFHSs FFHSs described described above above were were decreased decreased comparedcompared to to the the first first part part of of this this study: study: thethe totaltotal MA loading into into the the DEMO-FNS DEMO-FNS and and PIHR PIHR blankets blankets is 20is 20 t and t and into into IHR IHR is is 40 40 t. t. The The e efficiencyfficiency ofof thethe actinides’actinides’ incineration is is slightly slightly lower. lower. TheThe net net installed installed capacity capacity of nuclear of nuclear power power plants plants defines defines the total the SNF total (and SNF hence (and MA) hence production, MA) as wellproduction, as the totalas well accumulated as the total amountaccumulated of MA amount for a given of MA year. for a This given value year. defines This value the feasibility defines the of a fullfeasibility blanket loading. of a full blanket The results loading. presented The results here werepresented obtained here under were obtained the assumption under the that assumption the blankets werethat fully theloaded. blankets For were this, fully DEMO-FNS loaded. isFor decommissioned this, DEMO-FNS when is PIHRdecommiss is commissioned,ioned when whilePIHR PIHR is is decommissionedcommissioned, while when PIHR IHR is is decommi commissioned.ssioned when IHR is commissioned. FigureFigure 10 10 shows shows the the total total amount amount of of minorminor actinidesactinides in the the system system at at five-year five-year intervals intervals between between 19701970 and and 2130. 2130. 287 287 t oft of MA MA are are accumulated accumulated inin SNFSNF if there is is no no fuel fuel reprocessing, reprocessing, as as shown shown by by the the blueblue curve. curve. The The orange orange curve curve shows shows howhow thisthis massmass of actinides would would be be reduced reduced by by the the means means of of hybridhybrid reactors. reactors. The The third third curve curve describes describes thethe reductionreduction of MA MA by by the the hybrid hybrid reactors’ reactors’ operation operation and and the initial fuel loadings in the blankets, so it is possible to reduce the total amount by 28% (82 t) or the initial fuel loadings in the blankets, so it is possible to reduce the total amount by 28% (82 t) or 43% 43% (122 t), taking into account the initial fuel loadings. (122 t), taking into account the initial fuel loadings. Appl. Sci. 2020, 10, x FOR PEER REVIEW 13 of 16

FigureFigure 10.10. TotalTotal inventoryinventory of MA MA in in the the system. system.

The production rate of minor actinides in the system was obtained based on information about MA accumulation during the assumed period of time (Figure 11). As follows from the graph it will be necessary to incinerate more than 4 t/year of MA at the end of this timeline. This means that at least 4 facilities with IHR parameters will be needed.

Figure 11. Production rate of MA in the system.

Meanwhile, it is possible to take into commission just 2 additional IHRs in 2090 and 2120. In this case, it is possible to reduce the total amount by 48% (142 t) or 88% (262 t) taking into account the initial fuel loadings. It is not only the accumulated amount of minor actinides in SNF that is matter, but also the amount of minor actinides extracted in a given year during this timeline. The extracted amount is determined by the capabilities of Russia’s reprocessing plants. This determines the feasibility of full fuel loading in the blanket. 20 t of MA are needed for the blanket’s first loading. But only about 4 t of extracted MA at the start of DEMO-FNS in 2033, and 22 t in 2055, will be available. It is clear that there are not enough extracted minor actinides in the system when each hybrid reactor commences operations. The graph in Figure 12 shows the difference between available extracted amount of MA and needed amount for full blankets loading. If only the considered installed capacity of reprocessing plants is utilized, the deficit will not be overcome until after 2070. The negative values mean that the Appl. Sci. 2020, 10, x FOR PEER REVIEW 13 of 16

Appl. Sci. 2020, 10, 8417 13 of 15 Figure 10. Total inventory of MA in the system.

TheThe production production rate rate of of minor minor actinides actinides inin thethe systemsystem was obtained obtained based based on on information information about about MAMA accumulation accumulation during during the the assumed assumed period period of of time time (Figure (Figure 11 11).). As As follows follows fromfrom thethe graphgraph itit willwill be necessarybe necessary to incinerate to incinerate more more than 4than t/year 4 t/year of MA of atMA the at end theof end this of timeline. this timeline. This This means means that that at least at 4 facilitiesleast 4 with facilities IHR with parameters IHR parameters will be needed. will be needed.

FigureFigure 11. 11.Production Production raterate of MA in in the the system. system.

Meanwhile,Meanwhile, it isit possibleis possible to to take take into into commission commission justjust 2 additional IHRs IHRs in in 2090 2090 and and 2120. 2120. In Inthis this case,case, it is it possible is possible to reduce to reduce the the total total amount amount by 48%by 48% (142 (142 t) or t) 88% or 88% (262 (262 t) taking t) taking into into account account the initialthe fuelinitial loadings. fuel loadings. It isIt not is not only only the accumulatedthe accumulated amount amount of minor of minor actinides actinides in SNF in SNF that isthat matter, is matter, but also but thealso amount the of minoramount actinides of minor extracted actinides in extracted a given yearin a given during year this during timeline. this The timeline. extracted The amountextracted is amount determined is by thedetermined capabilities by the of Russia’scapabilities reprocessing of Russia’s plants. reprocessing This determines plants. This the determines feasibility the of fullfeasibility fuel loading of full in thefuel blanket. loading 20 tin of the MA blanket. are needed 20 t of for MA the are blanket’s needed fo firstr the loading. blanket’s But first only loading. about But 4 t ofonly extracted about 4 MAt of at theextracted start of DEMO-FNS MA at the start in 2033, of DEMO-FNS and 22 t in 2055,in 2033, will and be 22 available. t in 2055, It will is clear be available. that there It are is clear not enough that extractedthere are minor not enough actinides extracted in the system minor whenactinides each in hybrid the system reactor when commences each hybrid operations. reactor commences The graph in operations. The graph in Figure 12 shows the difference between available extracted amount of MA Appl.Figure Sci. 2020 12, shows10, x FOR the PEER diff erenceREVIEW between available extracted amount of MA and needed amount for14full of 16 and needed amount for full blankets loading. If only the considered installed capacity of reprocessing blankets loading. If only the considered installed capacity of reprocessing plants is utilized, the deficit plants is utilized, the deficit will not be overcome until after 2070. The negative values mean that the availablewill not amount be overcome of MA until in afterthe system 2070. The is less negative than valuesthe required mean thatamount the available for FFHSs’ amount operation of MA in in a particularthe system year. is less than the required amount for FFHSs’ operation in a particular year.

FigureFigure 12. 12.DifferenceDiﬀerence between between available available extracted extracted amount amount of MA of and MA needed and needed amount amount for full blankets for full loading.blankets loading.

4. Conclusions This research was performed as part of the Kurchatov Institute project on the development of fusion-fission hybrid reactors in Russia. This study shows that the application of fusion-fission hybrid reactors possesses a promising application potential for the transmutation of minor actinides. Calculations on neutron transport and nuclide kinetics were performed for particular reactors: demo, pilot industrial and industrial hybrid. It is shown that the capture to fission ratio is less than 1 for most actinides, which is a really positive feature for effective MA transmutation. The exceptions are Np-237 (only if H2O is used as a coolant), Am-241 and Am-243. For all of these, α is slightly higher than 1. Utilization of CO2 as a coolant leads to reduction in the total fuel loading, thus the potential advantages of this coolant for transmutation are neutralized. This situation could be improved via optimization of the transmutation area for this coolant. During irradiation, new actinides appear. This is an important feature for multicycle reprocessing and utilizing the fuel. Determining an equilibrium inventory for the FFHS’s fuel is an objective for future research as well as analysis of the inventory evolution and its influence on the neutron spectrum during irradiation and vice versa. For effective MA transmutation, in terms of not only technical but also economical parameters, it is necessary to use a FFHS with a large fuel loading and high capacity factor, such as IHR, which is able to incinerate more than 1 t of actinides per year and generate about 1 GW of electrical power. A system analysis of nuclear power in Russia, performed with the involvement of the hybrid reactors, highlights the problems associated with the absence of MA transmutation and the potential of FFHSs to reduce these hazards. As can be seen from the results obtained, there is a deficit of extracted MA, even for optimized development scenarios. At the same time, it is necessary to decommission the DEMO-FNS and PIHR (at least as burner reactors), and to abandon one additional IHR, which is needed due to the rate of MA accumulation in the system. It is also necessary to delay the commissioning of 2 other additional IHRs. For the considered scenarios, the deficit problem of extracted MA may be overcome around 2075. However, in the case of additional IHRs, the MA deficit takes place again at the moment when the second and third IHR are commissioned. This situation can be improved by the utilization of imported foreign SNF, which can be considered a service of radioactive waste management. The results obtained also indicate the need for faster commissioning the capacities of SNF reprocessing plants, to ensure the required amount of MA for FFHS fuel manufacturing. Appl. Sci. 2020, 10, 8417 14 of 15

4. Conclusions This research was performed as part of the Kurchatov Institute project on the development of fusion-fission hybrid reactors in Russia. This study shows that the application of fusion-fission hybrid reactors possesses a promising application potential for the transmutation of minor actinides. Calculations on neutron transport and nuclide kinetics were performed for particular reactors: demo, pilot industrial and industrial hybrid. It is shown that the capture to fission ratio is less than 1 for most actinides, which is a really positive feature for effective MA transmutation. The exceptions are Np-237 (only if H2O is used as a coolant), Am-241 and Am-243. For all of these, α is slightly higher than 1. Utilization of CO2 as a coolant leads to reduction in the total fuel loading, thus the potential advantages of this coolant for transmutation are neutralized. This situation could be improved via optimization of the transmutation area for this coolant. During irradiation, new actinides appear. This is an important feature for multicycle reprocessing and utilizing the fuel. Determining an equilibrium inventory for the FFHS’s fuel is an objective for future research as well as analysis of the inventory evolution and its influence on the neutron spectrum during irradiation and vice versa. For effective MA transmutation, in terms of not only technical but also economical parameters, it is necessary to use a FFHS with a large fuel loading and high capacity factor, such as IHR, which is able to incinerate more than 1 t of actinides per year and generate about 1 GW of electrical power. A system analysis of nuclear power in Russia, performed with the involvement of the hybrid reactors, highlights the problems associated with the absence of MA transmutation and the potential of FFHSs to reduce these hazards. As can be seen from the results obtained, there is a deficit of extracted MA, even for optimized development scenarios. At the same time, it is necessary to decommission the DEMO-FNS and PIHR (at least as burner reactors), and to abandon one additional IHR, which is needed due to the rate of MA accumulation in the system. It is also necessary to delay the commissioning of 2 other additional IHRs. For the considered scenarios, the deficit problem of extracted MA may be overcome around 2075. However, in the case of additional IHRs, the MA deficit takes place again at the moment when the second and third IHR are commissioned. This situation can be improved by the utilization of imported foreign SNF, which can be considered a service of radioactive waste management. The results obtained also indicate the need for faster commissioning the capacities of SNF reprocessing plants, to ensure the required amount of MA for FFHS fuel manufacturing. It follows from the analysis performed that, over the period of time considered, it is possible to decrease by 2130 the amount of MA in the system by ~28%, in the case of just 1 IHR, and by ~48%, in the case of two additional IHRs. If the MA loaded into the FFHS blankets are not considered to be waste, then the amount of MA excluded from the system may be increased by up ~43% and, in the case of two additional IHRs, by up to ~88%.

Author Contributions: Methodology, investigation, writing—original draft preparation and data curation, M.S.; supervision, conceptualization, writing—review & editing, B.K. All authors have read and agreed to the published version of the manuscript. Funding: This work was supported by contract № 1/17518-D/ between ROSATOM and NRC “Kurchatov Institute”. Acknowledgments: The authors appreciate the support from B. Kuteev as scientific supervisor, E.V. Muraviev and A.A. Kashirsky as consultants and the authors of the Universal System Model. Conflicts of Interest: The authors declare no conflict of interest.

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