NeutronicPerformance of theVVER-1000ReactorUsingThorium Fuel with ENDF Library

In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed usingMCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. (e effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. (e reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. (e calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff and the Doppler coefficient (DC) are found to be in agreement with the design values. It is found that the core loaded with uranium and thorium has lower delayed neutron fraction than the uranium oxide core.(emoderator temperature coefficients of the uranium-thorium core are found to be higher than those of the uranium core. Results indicated that thorium has lower production of minor actinides (MAs) and transuranic elements (mainly plutonium isotopes) compared with the relatively large amounts produced from the uranium-based fuel UO2.


Introduction
Countries are building nuclear power plants to meet their energy needs; the site of nuclear power plant in Egypt is El Dabaa. e VVER-1200 is the predecessor of the VVER-1000 reactor. e VVER-1000 is a 3000 MW thermal power nuclear power plant which is cooled and moderated by light water. e core is filled with 163 enriched uranium (UO 2 ) fuel assemblies (FAs). e reactor includes international safety standards with evolutionary design improvement in the areas of fuel technology, modularized construction, safety systems, and standardized designs [1,2].
In fact, thorium is three times more abundant than uranium, and the feasibility of loading the core with thorium uranium fuel was carried out. orium is made up mainly of "fertile" isotope ( 232 ). Experiments have been made on power reactors that were successfully operated using O 2 -UO 2 fuel in light water reactors (LWRs). 233 U and 232 are the best "fissile" and "fertile" materials, respectively, for thermal neutron reactors [3]. e neutronic parameter effects due to loading thorium as a part of VVER-1200 fuel under normal operation were studied by Dwiddar et al. [4]. e investigation used two different configurations, mixed uranium fuel with thorium and seed-blanket fuel. e amount of thorium inserted and the location of the thorium assemblies in the core of the reactor presented the master factor in determining the value of k eff and the core cycle length. e results concluded that the safest position of thorium is in the periphery of the core, and it is recommended to provide a blanket of mixed thorium uranium fuel for the entire core as the peripheral layer with the highest value of uranium enrichment. Mixing thorium with plutonium instead of uranium for the uranium enrichment, limited to 5%, was suggested.
Minimizing the fission products (FPs) using thorium as a fuel in place of the traditional UO 2 fuel has been studied by Galahom [5].
is paper discussed the VVER-1200 neutronic features for one assembly. e assembly fuel is a mix of thorium (fertile) with uranium (fissile). e neutronic features were calculated and compared with the traditional UO 2 fuel by using MCNPX code. e purpose of this study is to minimize the production of long-lived actinides and improve the fuel cycle length. e moderator temperature coefficient and the percentage of fissile inventory have been measured at different burn steps. e results showed that thorium fuels were better than UO 2 , and by replacing 238 U with 232 , we need much 235 U as a fissile material to sustain the same burnup level. For 232 + 235 U fuel, the Doppler coefficient (DC) was more negative [5]. Production of longlived actinides using thorium, fertile material, was reduced. e possible advantages of the seed-blanket (SB) assembly used in the VVER-1200 core instead of the homogeneous assembly were investigated using MCNPX 2.7 code and the ENDF/B-VII library [6]. e blanket region fertile material was 232 , while in the seed region, four different fissile materials have been investigated. e work concluded that in the homogeneous assembly, the power distribution is flatter than that of the heterogeneous assembly. e suggested fuels achieved a longer fuel cycle and a higher conversion ratio in the SB assembly than the homogeneous assembly. Moreover, using 232 instead of 238 U resulted in reducing the plutonium and the production of transuranic atoms.
Heterogeneous and homogeneous seed-blanket concepts were used to study the thermal hydraulics feasibility and the neutronics to change the AP1000 PWR reactor from UO 2 to (U-)O 2 [7]. e burnable poison geometry and materials in the core were not varied and only the fuel pin material was varied keeping the low enriched uranium (LEU) for 235 U. e study used three different fuel types, the first being 68% O 2 -32% UO 2 ; the second being 76% O 2 -24% UO 2 ; and the third being 80% O 2 -20% UO 2 . e work goals were to increase 233 U and minimize production of plutonium. e results showed that the homogeneous concept with three distinct types of fuel meets the optimization requirements. e results revealed that (U-)O2 hase many benefits, including a lower power density, retaining the same 18 months cycle, and a lower concentration of B-10 in the soluble poison and the elimination of B-10 in the coated integral boron poison [7].
In this paper, the neutronic performance and core analysis of the VVER-1000 one-six core were performed from the beginning of life (BOL) at cold zero power state (CZP) to the end of life (EOL) at hot full power state (HFP). e VVER-1000 reference core was initially filled with enriched uranium oxide fuel and then fueled with uraniumthorium fuel. Since thorium-based fuel is more favorable from the safeguard viewpoint, it is investigated in the present work as a fuel option to increase the fuel cycle length and reduce the plutonium amount produced. e neutronics calculations were made by MCNP6 code with two cross section libraries ENDF/B-VII.1 and ENDF/B-VIII.

The VVER-1000 Reactor Core Description
e VVER-1000 reactor type is a four-loop Russian version of the pressurized water reactor (PWR) producing about 1000 MW electric power. Figure 1 shows the core loading pattern for the first cycle of operation at the beginning of life (BOL) [8]. e reactor fresh core consists of FAs, which differ from one another by the enrichment of the fuel. e brief details of the core FAs were illustrated by Aghaie et al. [9]. e reactor core, FA, and fuel rod characteristics are shown in Table 1 [10].
In order to provide lower center temperatures and a free volume to allow any released fission gas to expand and thus decrease internal pressure, the fuel contains a central void hole in its fuel pellet [11]. e form of FAs, numbers, and average enrichment loaded for the first cycle of operation are shown in Table 2 [12]. e core configuration for the mixed thorium-uranium fuel used in this simulation is shown in Figure 2, where the 3 rd fuel batch with 4.95% enrichment is composed of 50% UO 2 and 50% O 2 .

Methods of Calculation and Validation
MCNP6 is the result of merging MCNP5 and MCNPX codes with new available options capabilities and features. MCNP6 includes more capabilities such as Shannon entropy, more options for tally treatments and tagging, variance reduction control methods, and more options for geometry and particles. is code is high fidelity due to its accuracy in simulating geometries and materials with continuous energy processing of nuclear parameters [13].
e largest update to the ENDF library is represented in ENDF/B-VIII. ese neutron sublibraries have been expanded by 32% to include 557 evaluations. ENDF/B-VIII and ENDF/B-VII.1 were the libraries used in this study [14,15].
In this study, the VVER-1000 one-six core shown in Figure 3 was simulated. It was fueled firstly with UO 2 and then fueled with O 2 +UO 2 . e neutronic parameters are presented for the two fuels. e simulation performed used 100000 neutron per cycle, 150 skipped cycles, and 250 active cycles.
ENDF/B-VIII library has the maximum neutron reactions changes on nuclides including actinides which affect nuclear criticality simulations. Two of the most important re-evaluated isotopes are uranium-235 and uranium-238. e re-evaluated parameters include the following.  [14,16] is illustrated in Table 3. (iv) (n,f ) Prompt fission neutron spectrum (PFNS): the ENDF/B-VIII evaluation for the PFNS mean energy is clearly softer than that of ENDF/B-VII.1, but it fits well with experimental data. e new average released neutron energy becomes 2.00 ± 0.01 MeV; on the other hand, it was 2.03 MeV in thermal range [16].
(v) Cross sections of (n, n') and (n, xn): the ENDF/B-VIII evaluation for the total inelastic scattering cross section (n, n') is slightly decreased than the last ENDF/B-VII.1 evaluation. ENDF/B-VIII valuation for the (n, xn) secondary neutron was not changed but showed a difference above 14 MeV. (vi) Nubar: evaluators used the parameter nubar to study criticality problems since criticality is highly sensitive to nubar. Several simulations have shown that the use of the new PFNS and TNCs of ENDF/B-VIII produces marginally higher k eff values than those of ENDF/B-VII.1 [17,18].   e neutronic parameters were calculated by MCNP6 with ENDF/B-VII.1 and ENDF/B-VIII and compared with the reactor design parameters [19] and published results [9,20]. It is necessary for any reactor core to have delayed neutrons since they can control the increase in the reactor power [21]. β eff is calculated by using the following equation:

Results and Discussion
where k prompt is the effective multiplication factor using only prompt neutrons, while k eff is the effective multiplication factor using both prompt and delayed neutrons. e fuel temperature coefficient of reactivity (DC) can be calculated by using the following equation: K 1 is calculated in case moderator and fuel are at 300 K and k 2 is calculated in case moderator is at 300 K while fuel is at 600 K [22]. MTC was determined by the following equation: K 3 is calculated in case fuel and moderator are at 600 K.

e Beginning of Life (BOL)
Results. e comparison between the calculated parameters with the published results is shown in Table 4. is comparison allows assessing the validity of the current VVER-1000 simulation. A general overall good agreement can be observed. In case of fueling the core with UO 2 , β eff values were 0.0078 and 0.00688 with ENDF/B-VII.1 and ENDF/B-VIII, respectively, while the expected value from the 235 U fission is 0.00650 according to Lamarsh and Baratta [23], so it was realistic to see the core has a delayed neutron fraction that is greater than this expected value. According to RFANE [19], the delayed neutron fraction is 0.00740 and 0.007110 according to Gholamzadeh et al. [20] which is in agreement with the obtained results. e fraction of the delayed neutrons differs from fuel material to another. β eff was 0.0031 and 0.0069 for 233 U and 235 U materials, respectively, according to IAEA [3] which means that the O 2 +UO 2 core is expected to have a low delayed neutron fraction and this is clear in Table 4. β eff values are 0.00628 and 0.00545 with ENDF/B-VII.1 and ENDF/B-VIII falling in the expected range for O 2 +UO 2 core.

Temperature Coefficients of Reactivity.
is section discusses the reactor operation safety parameters which are the variation of the reactivity with temperatures (the temperature coefficients of reactivity). Table 4 presents the main operating safety parameters (DC and MTC). As the temperature of the moderator increases, its density decreases and lower fraction of neutrons is slowed down, resulting in negative change of reactivity. On the other hand, the neutron spectrum hardens due to the decrease of moderator fraction, which will decrease the core reactivity. e above two factors compete with each other and might lead to a negative MTC. e DC is the most important safety parameter because it measures the reactor operation stability. e higher temperature of fuel permits the fertile material to absorb many neutrons away from the fission. 232 absorbs larger neutrons than 238 U in higher fuel temperatures. erefore, thorium-based fuel has more negative DC than the uranium fuels; this is clearly shown in Table 4. e MTC is the change in reactivity by changing the moderator temperature. e reactor is designed to have negative MTC value to provide negative reactivity feedback, which indicates that the more the temperature increases, the more the reactivity decreases. e UO 2 MTC is more negative than O 2 +UO 2 ; this is because of the larger fast fission cross section of U-238 than that of -232, and this agrees with the results in Table 3. e higher temperature of fuel permits the fertile material to

e Middle of Life (MOL)
Results. e behavior of k eff over time for the VVER-1000 core can be seen in Figure 4. e calculations were performed along with 5 burnup steps of 100 days. e standard deviations for k eff values ranged from 0.00036 to 0.0004 for UO 2 fuel and from 0.00036 to 0.00041 for O 2 +UO 2 fuel. k eff values were 1.19911 and 1.23126 for UO 2 and O 2 +UO 2 , respectively, at BOL. It then decreased with the depletion of the fuel. O 2 +UO 2 core can keep criticality for more than 500 effective full power days while the UO 2 core can keep criticality nearly up to 450 effective full power days. In the following figures, UO 2 -7 represents current results for UO 2 core by ENDF/B-VII.1; UO 2 -8 represents current results for UO 2 core by ENDF/B-VIII; U -7 represents current results for O 2 +UO 2 core by ENDF/B-VII.1; U -8 represents current results for O 2 +UO 2 core by ENDF/B-VIII.
As mentioned before, this simulation uses the O 2 +UO 2 core configuration where the 3 rd fuel batch with 4.95% enrichment is composed of 50% UO 2 and 50% O 2 . Not only the thorium amount inserted but also the thorium assemblies location in the reactor play the principal role in k eff value determination and the core cycle length. e thorium location in the periphery results in higher value of k eff and longer length of cycle [3].

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After 500 effective full power days, the O 2 +UO 2 core remains supercritical and this can be explained by the fact that since only half of the UO 2 fuel is replaced with O 2 which is located in the low neutron flux periphery assemblies. Moreover, insertion of thorium increases the first core cycle length compared with UO 2 core fuel cycle, and increasing the UO 2 enrichment may help this. For this reason, in order to obtain more optimum advantage of thorium, it has to be rearranged in the middle of the core. is strategy can be realized by locating O 2 in the core periphery for the first operational cycle and then shuffling them into the core interior in the subsequent cycles. orium was mixed with 235 U to provide power till building enough 233 U amounts. Figure 5 illustrates the change in the burnup rate over time. As time passes, the burnup rate increases till it reaches 20.13 and 21.3 GWD/ MTU for the UO 2 and O 2 +UO 2 one-six core, respectively, at end of life (EOL). Figure 5 shows that the core burnup results of ENDF/B-VII.1library are typically the results of the ENDF/B-VIII library, and the changes appear only in the criticality calculations.

e End of Life (EOL) Results.
e 235 U is the core fissionable material in that it steadily decreases as it burns over the lifetime of the core, in addition to increasing the output of 239 Pu and 240 Pu levels towards the EOL. e fissile component 235 U is depleted from the startup until EOL, and other fissile components are created when 238 U is transmuted to higher actinides, especially 239 Pu and 240 Pu. e quantity of total fission products (FPs) in the core grows up as the burnup level increases. FPs are neutron absorbers which have a strong adverse effect on the neutron economy of the core as time passes. Figures 6 and 7 show the fuel consumption and the production of actinides that build up in the UO 2 and O 2 +UO 2 cores. e UO 2 core contains 235 U+ 238 U; the two isotopes were re-evaluated in the ENDF/B-VIII library which means marginal differences in results, and this appears in Figure 6(a) especially for the consumption of 235 U. On the other hand, ENDF/B-VIII did not re-evaluate thorium. e   Science and Technology of Nuclear Installations thorium core contains 235 U+ 238 U+ 232 , hence the slight difference in the U -7 and U -8 mainly from 235 U+ 238 U cross section difference. e principal benefit of using thorium as a fuel option is to get as much fissile isotope 233 U as possible and eliminate the plutonium production, and this is clearly demonstrated in Figure 7. Figure 8 shows the neptunium isotope masses produced during the reactor operational cycle. Figure 9 presents the consumption of thorium in the O2+UO2 core. ere was a thorium consumption difference between ENDF/B-VII.1 and ENDF/B-VIII at 200 days, but we do not have any explanation for this difference.

Conclusion
From the beginning of life to the end of life, the VVER-1000 neutronic parameters were calculated using MCNP6 code. ese parameters were determined for a UO 2 and a mixed O 2 +UO 2 fuel. e obtained results fall within an acceptable range with respect to the reactor design parameters and the published data. e delayed neutron fraction β eff values were found to be 0.0078 and 0.00688 for UO 2 fuel, with ENDF/B-VII.1 and ENDF/B-VIII, respectively, which are in an agreement with the reference value of 0.00740. On the other hand, O 2 +UO 2 has lower delayed neutron fraction β eff values (0.00628 and 0.00545) than UO 2 fuel. For the UO 2 fuel, its MTC value was −1.39 × 10 −4 1 (°C) while for the O 2 +UO 2 fuel, it was −0.339 × 10 −4 (°C) with ENDF/B-VII.1. All fall in the reactor design operational limits. It has been shown that the MTC of the UO 2 core is more negative than that of the O 2 +UO 2 one; this is because of the larger fast fission cross section of 238 U than that of 232 . oriumbased fuel has more negative DC than the uranium fuels. e behavior of k eff over time for both cores explained the slight increment of ENDF/B-VIII than ENDF/B-VII.1 in the criticality calculations for the UO 2 fuel. k eff decreased with the depletion of the fuel indicating that the UO 2 core has to be refueled earlier than the O 2 +UO 2 core for the first cycle of operation. e plutonium isotopes produced due to the O 2 +UO 2 core were lower than those of the UO 2 core.

Data Availability
e data used to support the findings of this study are included within the article.

Conflicts of Interest
e authors declare that they have no conflicts of interest.   Science and Technology of Nuclear Installations