Overview of Fourth Generation Reactors Studied at the Universidade Federal de Minas Gerais (Brazil)

: This paper presents the results of the studies about the Fourth Generation Reactors developed at DENU-UFMG, such as the Very High Temperature Reactor (VHTR), Molten Salt Breeder Reactor (MSBR), Sodium Fast Reactor (SFR) and Gas-cooled Fast Reactor (GFR). These studies evaluate neutronic parameters of the nuclear system at steady state and during the burnup using computer models developed at DENU-UFMG using SCALE 6.0 (KENO-VI/ORIGENS), MCNPX 2.6.0 and MCNP5. The results show that the adopted methodologies confirm the capabilities of the codes to simulate specific situations in steady state or transient operating conditions.


INTRODUCTION
The demand projections of the global energy show an increasing trend and require greater use of reliable energy sources.Nuclear power plants may represent very attractive options for energy generation because of its high compliance degree with sustainability criteria and its high capacity factor compared to other energy alternatives [1].However, there are still economics, proliferation, safety, and waste production issues that innovative approaches should enhance.The new generation of nuclear systems presents advanced concepts that may break out of the limitations of currently used nuclear energy systems.The Fourth Generation Reactors offer potential benefits, including possible lower costs, enhanced safety, greater resource utilization, waste minimization, co-production of process heat for industrial operations, improved proliferation resistance, and easier operation [2].
In the last decades, several researches have been developed around the world where the "International Project on Innovative Nuclear Reactors and Fuel Cycles" (INPRO) and the "Generation IV Forum" (GIF) are important initiatives concerning the discussions on the future of nuclear reactor technology.Formally, Brazil participates of INPRO and was also involved in GIF until recently.For better efforts coordination of Brazilian researchers and aiming at increase the importance and visibility of Brazil in the international forums on innovative and advanced nuclear systems, the "National Institute for Advanced and Innovative Nuclear Reactors (INCT-RNI)" was created in 2008 [3].This institute has as mission to perform research and prepare human resources for developing innovative nuclear reactor technologies for the sustainable development of Brazil.
Among the scientific institutions that compose the INCT-RNI is the Nuclear Engineering Department of the Universidade Federal de Minas Gerais (DENU-UFMG), which has been studying advanced nuclear systems.The research group of DENU-UFMG p. 4 has as a goal evaluates the innovative fuel cycle associated with neutronic analysis of the new reactor generation, as it is presented in this paper.The reactors VHTR, MSBR, GFR, and SFR have been simulated by nuclear codes as SCALE 6.0 (KENO-VI/ORIGENS), MCNPX 2.6.0 and MCNP5 (e.g. ).These nuclear systems will tend to have closed fuel cycles and transmute the long-lived actinides: they are expected to have full actinide recycling.
In this way, the present paper overviews the simulated systems, and the results are discussed as follows.

USED CODES
Monte Carlo (MC) codes have been widely used around the world to simulate the neutronic behavior of different nuclear reactors.The SCALE 6.0 and MCNPX 2.6.0 are mostly MC codes used to analyze steady-state neutronic parameters such as the multiplication factor, neutron flux, reaction rate, reactor physics constants, and time dependent burnup calculations.The steady-state analyses are performed by particle transport calculus using MC method.This technique plays a relevant role to perform the neutronic calculation over the spatial dependence of the neutron flux in fully three-dimensional geometry.On the other hand, the time dependent calculation is a linked process that involves the steady-state results with the nuclide depletion that is performed by an external module, for example, SCALE 6.0 uses KENO-VI coupling with ORIGENS to burnup calculations, while MCNPX 2.6.0 uses MCNP coupling with CINDER-90 module.These coupled calculations aim to predict the temporal changes in the core's material composition and neutronic parameters changes.However, stochastic codes present a larger computational cost than deterministic methods, mainly due to geometry details.This characteristic implicates to the enormous computational time in the burnup calculation.In this work, the steady-state simulations are performed before the time dependent calculation to optimize the computational time.This methodology was employed in MHR, GFR and SFR systems.In p. 5 the MSBR, the coolant/fuel (molten salt) flows through the reactor core and it is subjected to continuous reprocessing.The MCNPX 2.6.0 and SCALE 6.0 do not comprise this continuous reprocessing in the fuel depletion calculation which is different of other types of nuclear systems.Thus, the MSBR study presents analysis only at steady-state aims and evaluates the fuel composition for future burnup calculation studies.
The MCNPX uses this library as a continuous one, and the SCALE6.0code uses 238 groups collapsed from the ENDF-B-VII.0 continuous library (V7-238).

Very High Temperature Reactor
Gas-cooled reactors have been investigated since the early days of nuclear power.The early gas reactors were used commercially in the United Kingdom [27] but were overshadowed by the Light Water Reactors (LWR) and other designs.Recently, interest in gas-cooled reactors has been renewed, and several High Temperature Gas Reactor (HTGR) designs have been developed for near-term deployment.The VHTR concept is an evolutionary extension beyond the near term HTGR designs.The VHTR is a helium-cooled, graphite-moderated reactor with a ceramic core and TRISO coated fuel particles, where the fissile fuel is enclosed within layers of temperature resistant materials, including SiC and pyrolytic graphite.These particles are compacted with graphite powder to form the fuel pins inserted into hexagonal fuel assemblies.The reactor core consists of an array of hexagonal graphite blocks distributed in three regions: inner ring, central ring and outer ring.The reflector blocks are placed into inner and external rings.The fuel assemblies and the control rods blocks are arranged into the central ring [28].This geometry was modeled in the MCNP5 code (Figure 1) where the core design was based on previous studies [29].Firstly, two models were evaluated to verify the geometry sensitivity in the simulations: heterogeneous (HE) and semi-homogeneous (SH).The HE model simulates the TRISO particles geometries and compositions.In the SH model, the TRISOs were not simulated but the particles densities and the compositions were homogenized into fuel pins.Table 1 presents the effective multiplication factor (keff) of the VHTR core for HE and SH models and for the reference case (RC) [29] with UO2 fuel enriched at 15%. Figure 2 depicts the MCNP5 modeling for both the HE and SH cases.Among the cases, the results of RC and SH have the most considerable similarity.The difference of keff between them is 61.0 pcm, and considering the standard deviation, the keff of these models is statistically equal.On the other hand, between RC and HE model, this difference is 719 pcm.Thus, the SH model has a more significant agreement with the RC than the HE model because the SH model uses the same homogenization methodology than RC case.Despite of the differences, the models have similar keff values and to save computational time, the SH model can be used in the neutronic studies.
In VHTR concept, it is inevitable that the quantity of TRISO particles into fuel pins  p. 9 Considering that the fission probability to fissile isotopes is higher to thermal neutrons, a hardening neutron spectrum may reduce the fission reaction number.This behavior may contribute to stabilizing keff values as PF increases.This tendency is more remarkable in the highest FC values (Figure 3 and 4), because of the increase in the hardening neutron spectrum of these cases (Figure 5 and 6).The keff and φF/φT present similar values to FC = 13.0,16.0 and 20.0, with the smallest difference being PF ≥ 0.30.
Between the two fuel types, for the same PF-FC value, the U-Pu has smallest keff and the most hardened neutron spectrum.This behavior may be due to Pu isotopes, which have high probability of radiative capture reaction for thermal neutrons.
Based on reference value, the burnup study considers FC=16.0 among the evaluated cases.As verified, the smallest and the highest FC values generate the smallest and the highest keff respectively.Thus, the fuel depletion analysis regard to intermediate PF value (0.30).The simulations consider 1,000 days at hot full power with specific power density of 103 W/g.Between the two fuel types, the mass of 238 Pu, 239 Pu and 240 Pu reduce to U-Pu but increase to EU.Among these nuclides, the more expressive difference is to 239 Pu, were the U-Pu fuel presents a reduction of 8.8% and EU increase of 0.8% in 239 Pu mass (Figure 8).
This behavior is due to the fission of 239Pu in U-Pu and the transmutation of 238 U in 239 Pu p. 13 in EU.Also, the mass of Np is smallest to U-Pu fuel.Among the Np isotopes, the 237 Np presents the biggest difference (Figure 9) where the production this nuclide is higher to EU fuel.The 237 Np is formed through the consecutives reactions 235 U(n,γ)→ 236 U (n,γ)→ 237 U (β -)→ 237 Np.
Thus, as the EU has higher 235 U mass at BOL (Table 2), the concentration of 237Np is higher to this fuel type.
However, the U-Pu presents an increase of the 243 Pu, 244 Pu, Am and Cm, while the EU do not have these isotopes in its final composition (Table 2).Among these nuclides, the increase of 241 Am, 243 Am, 244 Cm and 242 Cm are more evident (Figure 9).This behavior may be due the presence of Pu in the U-Pu fuel where the beta decay of 241 Pu and 243 Pu produce 241 Am and 243 Am, respectively.Also, during the reactor cycle there is production 242 Am and 244 Am, which decay to 242 Cm and 244 Cm, respectively.
Therefore, concerning to reprocessed U-Pu fuel, the simulations present a reduction of Pu mass but an increase of Np, Am and Cm mass (Table 3) where the reduction of Pu is higher than the production of Np, Am, and Cm.

Molten Salt Breeder Reactor
This reactor concept was developed in the 1960s by Oak Ridge National Laboratory (ORNL), which was operated successfully demonstrated high burnup and on-line fueling [30].Today the Molten Salt Reactor is one of the promising future nuclear reactor concepts included in the Generation IV roadmap.In this nuclear system, the fuel is a circulating p. 14 mixture of salts between the reactor core and the heat exchanger.Thorium fueled MSR concept is one of the most promising nuclear reactors designs currently being studied because of the high natural availability of this fuel.The aim is use 232 Th to breed 233 U, where uranium ( 233 U or 235 U) is the initial source of fissile material that needs to be provided.
However, plutonium (Pu) and minor actinides (MAs) could be used instead of uranium for initial fissile loading.In this sense, the MCNPX 2.6.0 code was used to calculate the criticality and the neutron flux profile at the steady state of MSBR core.The MSBR configurations use the data from a conceptual design developed by ORNL [31].Figure 10 illustrates the geometry of the simulated system, where graphite is the main material other than salt which is used for neutron moderation and reflection.The reactor core has a square lattice that contains graphite blocks with a square cross section and a circular fuel channel.Firstly, the MCNPX model was compared to previous studies considering the conventional 233 U ( 232 Th fuel with 1.76% fissile content [32,33].Table 4 presents the p. 15 effective multiplication factor (keff) at beginning of cycle (BOC) of the MSBR system simulated by DENU-UFMG (this work), University of Illinois at Urbana-Champaign (UIUC) [32], Ulsan National Institute of Science and Technology (UNIST) [33] and Korea Atomic Energy Research Institute (KAERI) [33].In the present study, the calculated keff by the MCNPX 2.6.0 code agrees with others works.The highest difference is related to the SERPENT code.According to reference [32], this discrepancy may be due to the simplifications in Zone II performed in the SERPENT model.In the second step, different nuclides were evaluated as an initial seed of the system in the fission/transmutation process.Thus, the following fuel types were simulated: 1) Recovery Uranium (RU): Conventional 233 U from the U-Th cycle; 2) Enriched Uranium (EU): 235 U manufactured by enrichment plant; 3) Reprocessed Plutonium (RP): Nuclides of Pu from spent PWR fuel; and 4) Reprocessed Actinides (RA): Matrix of Pu/AMs from spent PWR fuel.
Table 5 presents the keff of the MSR core for the evaluated fuels.These fuels were simulated with two fissile isotopes concentrations because the ORNL reports two values which represent a conceptual (2.26%) and an optimized (1.76%) MSBR design [31].The results show that the same weight fraction of fissile isotopes does not produce similar keff values.This behavior can be due to: (a) the presence of nuclides in RP and RA fuels that have high neutrons cross section for radiative capture; and/or (b) the different atomic masses of the fissile isotopes, because the fuel that contains fissile isotopes with the lightest mass p. 16 has more fissile atoms.In this sense, the fissions number may be highest for the RU, producing highest keff value.Thus, increasing the concentration of fissile isotopes of RP and RA fuels is necessary, such that they have keff values similar to the RU fuel.

Gas Cooled Fast Reactor
The first fast breeder reactor, called Clementine, was constructed in the 1946 during the Manhattan Project at Los Alamos National Laboratory in the US.In the 1960s, prototype reactors operated in China, France, Germany, India, Japan, the Russian Federation, the p. 18 United Kingdom and the United States [34].Although there has been stagnation in the development of fast reactors in past decades, in the countries involved, this reactor concept represents an important role in the advanced reactors.In 2002, the Generation IV International Forum (GIF) selected six nuclear system concepts to compound the fourth generation reactors list.In this role, four types are fast reactors, among which is the Gascooled Fast Reactor (GFR).The GFR can fully utilize fuel resources, minimize or reduce its own (and other systems) actinide inventory, produce high efficiency electricity, and utilize high temperature process heat [35].In this way, the research group DENU-UFMG has been used the SCALE 6.0 and MCNPX 2.6.0 codes to evaluate the GFR criticality under steady state and burnup conditions.Three carbide fuel types were evaluated: 1) Natural uranium and plutonium -(U, Pu)C; 2) Depleted uranium and transuranic -(U,TRU)C; and 3) Thorium and transuranic -(Th, TRU)C Figure 12 shows the configured model comprising the active core without the reflector layers.Thus, to minimize the lack of reflector layers, these simulations comprise reflective surface boundaries, so there are no neutron leakages.The core dimensions, geometry and properties were based in the reference work [36].Previous studies [14,15] verified that the most suitable percent of the fissile material for each nuclear fuel is 11.33% for (U,Pu)C, 13.06% for (U,TRU)C and 15.78% for (Th,TRU)C which produce similar kinf for the three fuels.

p. 19
Figure 12: GFR core configuration  and End of Cycle (EOC) and the Figure13 illustrates the behavior of this factor during the burnup.In the simulations, the specific power of the GFR is 48 W/g.Between the codes, the kinf values are very similar and the small differences may be attributed to the data treatment between the codes.The MCNPX code uses a continuous library, but the SCALE6.0code uses 238 groups collapsed from the ENDF-B-VII.0.Also, MCNPX performs the burnup using the CINDER90 code using 63 energy groups and the SCALE6.0uses the ORIGEN-S, which performs the burnup using 3 energy groups.Among the fuels, the (Th, TRU)C has the biggest kinf at EOC, because of 233 U build-up effects.The (U,Pu)C and (U, TRU)C present similar kinf at EOC. presents an increase of 239 Pu and 241 Pu, and a decrease of 238 Pu (Figure 14).These behaviors can be due to the fission of 239 Pu and 241 Pu and because of the transmutation of 238 U in 238 Pu.
In the (U,TRU)C fuel, there is an increase of 244 Cm while the 241 Am and 237 Np concentrations are practically constant (Figure 15).Regard to (U,Th)C fuel, due to the 232 Th transmutation, there is a buildup of 233 Th and 233 Pa, which are responsible for the 233 U production (Figure 16).During the burnup, there is a gradual increase of 233 U, but it tends to stabilize during the reactor cycle.Among the evaluated fuels, the (Th,TRU)C presents the higher generation of fissile isotopes that produce the highest kinf at EOC (Figure 15).

Sodium Fast Reactor
The sodium-cooled fast reactor (SFR) is one of the four fast neutron concepts selected by the GIF withsignificant industrial experience feedback.Approximately, twenty prototypes or demonstrators have been built throughout the world, which has demonstrated the excellent use of the uranium resource, as well as, the capability of these reactors to recycle the plutonium without any limitation in the number of recycling operations (multi-recycling) [37].On the other hand, there is interest in developing small and medium sized reactors because of their attractive features as compatibility for small electricity grids; lower capital costs; reduced design complexity; among others advantages [38].Therefore, engaging these factors, the research group DENU-UFMG has studied a small fast sodium reactor concept similar to the 4S concept from Toshiba.This reactor includes passive nuclear safety systems and does not require on-site refueling.This concept is not a breeder reactor because there is no uranium blanket around the core for the absorption of neutrons leakage, but its hardening neutron spectrum provides higher power densities and higher fuel burnup.Figure 17 p.23 illustrates the SFR model simulated in MCNPX 2.6.0, and Table 8 presents the reactor core's enrichment percentage and fuel mass [39,40].Comparing Figure 19 and 20, the 235 U mass decrease is 70,9% and 75,3% for outer and inner zones, respectively.For the 238 U these reductions are 10,6% and 13,9%.In this way, the inner zone produces a bigger uranium reduction than the outer zone.However, the mass increase of the 239 Pu has an inverse behavior, that is, the outer zone presents the biggest mass of this nuclide (Figure 20).Despite 239 Pu production, it is possible that the inner region burns more 239 Pu by fission reaction than the outer zone.
p. 26  In addition to scientific studies, the developed models are used for academic studies.
The used codes are suitable for graduate students researching scientific calculations to master and/or doctoral thesis.Also, undergraduate students are encouraged to participate in research groups by specific projects of scientific education.All developed projects aim to produce scientific results and qualify professionals for the nuclear field.In this way, the Federal University of Minas Gerais has contributed to scientific research and the diffusion of nuclear knowledge.
has a fundamental role in the reactor performance.Different packing fraction (PF) of TRISO particles associated with different fissile content (FC) in the fuel, generates unlike behavior of neutronic parameters.In order to evaluate keff and neutron flux as functions of PF and FC, several PF-FC combinations were simulated using the SH model.Two fuel types were simulated: (a) enriched uranium (EU), and (b) a mix of depleted uranium and reprocessed plutonium (U-Pu).Figures3 and 4illustrate the keff values to EU and U-Pu respectively.In these figures, each curve represents a FC percentage in the fuel, where keff values increase as FC rises.Also, for a same FC value, the keff values increase as PF rises because the fuel volume is proportional to PF.As PF rises, the volume of fissile material increases and contributes to keff rise.As the total volume of fuel pin does not change, this PF increasing represents an increase in the TRISO particle volume, but a reduction in the graphite volume (moderator), which generates a hardening in the neutron flux.

Figure 3 :Figure 4 :
Figure 3: Effective multiplication factor as function PF-FE to EU fuel type

Figure 5 :Figure 6 :
Figure 5: Fast to total neutron flux ratio into fuel pin (EU fuel type)

Figure 7
Figure 7 depicts keff values during the irradiation cycle of VHTR to EU and U-Pu fuels.

Figure 7 :
Figure 7: Effective multiplication factor as a function of burnup

Figure 8 :Figure 9 :
Figure 8: Variation in U and Pu mass between BOL and EOL

Figure 11
Figure11illustrates the radial neutron flux profile in the reactor core to RU, RP, and RA fuels.The central region (Zone I) has the highest flux, while the outer zone presents the smallest one.This behavior is expected because the simulations do not comprise the reactivity control.The peripheral region is the reflector where the neutron flux is almost zero, so there is no neutron leakage from the reactor core.For RU fuel type, the center core presents a depletion of this flux which the absence of control rods may have produced.The withdrawal of the graphite control rod inserts negative reactivity to the core due to the decrease of neutron moderation.The reprocessed fuels (RP and RA) present a neutron flux profile more flatten than RA.Although the RP and RA have a concentration of fissile isotopes higher than RU, the reprocessed fuels have neutron absorbers that may cause a neutron flux reduction.This behavior suggests that to RP and RA, a different radial distribution of the fuel blocks into MSR core is necessary to flatten the neutron flux profile.

Figure 11 :
Figure 11: Total neutron flux in the MSR core

Figure 13 :
Figure 13: Infinite multiplication factor as a function of time

Figure 18 illustrates 25 Figure 18 :
Figure 18 illustrates the keff behavior during the burnup.The calculations were performed in two schemes, at the standard power of 135 MWth (Cycle 01) and another one of 215 MWth (Cycle 02) with the graphite reflector on the top of the core.Also, in the Cycle 02, the fixed absorbers are removed to allow positive core reactivity.The abrupt increase in the curve represents the EOC of Cycle 01 and the start of the BOC of Cycle 02.The keff value is 1.02775 and 1.02846 at BOC for Cycles 01 and 02, respectively.The reactor became sub-critical after 6,935 EFPDs (19 years) and 4,015 EFPDs (11 years) for Cycle 01 and Cycle 02, respectively.This results in a total of 10,950 EFPDs or 31 years of reactor operation.The keff in cycle 2 shows a curve with saw-toothed appearance between 25 to 31 years, which is caused by the computational effort to perform the calculation following changes in the fuel isotopes composition.

Figure 19 :Figure 20 :
Figure 19: Initial mass of the uranium for Cycle 01 at BOC

Figure 21 :
Figure 21: Neutron flux in three different reactor zones

Table 1 :
The keff of Reference Case (RC) and simulated models SH and HE

Table 2
presents the isotopic mass of the evaluated fuels at Beginning of Life (BOL), and End Of Life (EOL), and Figures8 and 9illustrate the mass variation (∆m (%)) between BOL and EOL, which was calculated regarding to total fuel mass.

Table 2 :
Isotopic mass (g) and mass variation (∆m) between BOL and EOL 1212

Table 3 :
Total mass variation (∆m(%)) between BOL and EOL by chemical element type

Table 4 :
MSBR criticality at BOC to RU fuel calculated by different institutes

Table 5 :
Effective multiplication factor for the evaluated fuels

Table 6
presents keff as a function of fissile isotopes content and heavy metal concentration.As expected, keff increases as fissile isotopes augment.The keff of RP and RA are similar to RU when the mass of the fissile isotopes is 16 and 25%, respectively.Then, this percentage was used to calculate the total neutron flux in MSR core.

Table 6 :
Effective multiplication factor as a function of fissile isotopes 1717

Table 7 :
Infinite multiplication factor at BOC and EOC considering 1400 days of burnup

Table 7
presents the infinite multiplication factor (kinf) at Beginning of Cycle (BOC)

Table 8 :
Basic parameters of the core model The design of the fuel is based on U-10Zr (10%w Zr) binary alloy, with HT-9 steel fuel rod cladding.The graphite reflectors can move axially to allow the reactivity control, which compensates fuel burnup and reduces neutron leakage.The burnup was calculated with a standard power operation of 135 MWth and considered an extended power operation of 215 MWth.The simulation takes the prediction of cycle extension of the 10,950 equivalent full power days (EFPDs) or about 31 years for extended power, and 16,790 EFPDs or about 47 years for standard power.