Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO₂, ThC, UO₂, UC, U₃Si₂ and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of ²³⁹Pu breeding capability among the uranium fuels whereas UO₂ has the lowest one. ²³⁹Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and ²³³U production values are 0.125 and 0.140 in the blanket fueled with ThO₂ and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm³ in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.     

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li₂₀Sn₈₀ coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.     

Numerical Study on Effects of Fuel Mixture Fraction and Li-6 Enrichment on Neutronic Parameters of a Fusion–Fission Hybrid Reactor

This study presents the effects of mixture fractions of nuclear fuels (mixture of fissile–fertile fuels and mixture of two different fertile fuels) and ⁶Li enrichment on the neutronic parameters (the tritium breeding ratio, TBR, the fission rate, FR, the energy multiplication ratio, M, the fissile breeding rate, FBR, the neutron leakage out of blanket, L, and the peak-to-average fission power density ratio, Γ) of a deuterium–tritium (D–T) fusion neutron-driven hybrid blanket. Three different fertile fuels (²³²Th, ²³⁸U and ²⁴⁴Cm), and one fissile fuel (²³⁵U) were selected as the nuclear fuel. Two different coolants (pressurized helium and natural lithium) were used for the nuclear heat transfer out of the fuel zone (FZ). The Boltzmann transport equation was solved numerically for obtaining the neutronic parameters with the help of the neutron transport code XSDRNPM/SCALE4.4a. In addition, these calculations were performed by also using the MCNP4B code. The sub-limits of the mixture fractions and ⁶Li enrichment were determined for the tritium self-sufficiency. The considered hybrid reactor can be operated in a self-sufficiency mode in the cases with the fuel mixtures mixed with a fraction of equal to or greater than these sub-limits. Furthermore, the numerical results show that the fissile fuel breeding and fission potentials of the blankets with the helium coolant are higher than with the lithium coolant.     

Oxidation of accident tolerant fuel candidates

In this study, the oxidation of various accident tolerant fuel candidates produced under different conditions have been evaluated and compared relative to the reference standard – UO₂. The candidates considered in this study were UN, U₃Si₂, U₃Si₅, and a composite material composed of UN–U₃Si₂. With the spark plasma sintering (SPS) method, it was possible to fabricate samples of UN with varying porosity, as well as a high-density composite of UN–U₃Si₂?(10%). Using thermogravimetry in air, the oxidation behaviors of each material and the various microstructures of UN were assessed. These results reveal that it is possible to fabricate UN to very high densities using the SPS method, such that its resistance to oxidation can be improved compared to U₃Si₅ and UO₂, and compete favorably with the principal ATF candidates, U₃Si₂, which shows a particularly violent reaction under the conditions of this study, and the UN–U₃Si₂?(10%) composite.     

Neutronic and thermal-hydraulic analysis of alternative ceramic fuels in the next-generation of light water reactors

The Burn-Up enlargement is one of the most important issues in the nuclear reactor core fuel management. In recent years some reactor design companies have focused on the reactor cycle length enlargement in next generation of pressurized water reactors. An increased cycle length results in an increased fuel burn-up which directly leads to low electricity costs and more efficiency. One of the promising issues is to change the chemical state of fuel that is on the agenda of the Mitsubishi Company as US-APWR nuclear power plant designer. In the present study, the neutronic as well as thermal-hydraulic analysis of some commercial ceramic fuels such as UN, UC, and UN¹⁵ instead of conventional UO₂ have been studied. The sub-channel analysis approach has been selected for these investigations. In this regard, a US-APWR fuel assembly was modelled using MCNPX2.6 Monte Carlo code by considering the periodic boundary condition in X–Y directions. It was found that the use of UC and UN¹⁵ instead of UO₂ has a deep effect on the reactor cycle length such that the power plant operational time was increased by a factor of 1.5. The COBRA-EN code with modified MATPRO subroutine has been used in thermal-hydraulic tasks. Since the thermal conductivity of these selected fuels is six times greater than UO₂, the thermal-hydraulic analysis of candidate fuels was led to outstanding results. It was found that the fuel centerline temperature in UN¹⁵ and UC cases are about half of UO₂ one, which is drastically beneficial. In summary the thermal power of next generation of pressurized water reactors could be increased considerably by using the candidate ceramic fuels instead of conventional UO₂ one.     

Impact of Resonance Treatment on Minor Actinide Incineration in a D–T Thorium Fusion Concept

Resonance treatments have an essential role to reliable neutronic calculations with different neutronic parameters. In this study presents the effect of resonance treatment and various tritium breeder materials on the incineration of the nitride fuels containing minor actinide mixed thoria in the Deuterium–Tritium fusion driven hybrid reactor as time dependent. Neutron transport calculations under resonance treatment and without resonance treatment are performed by using XSDRNPM/SCALE 5 codes. The impact of resonance treatments and various tritium breeder materials on tritium breeding, energy multiplication, total fission rate (∑_f), cumulative fissile fuel enrichment, fissile fuel breeding, average burn up values are comparatively investigated. It is observed that the neutronic results affect from both resonance treatment and the tritium breeder materials as time dependent.     

On the Neutronic Performance of Hylife-II Reactor Fuelled with Carbide Fuels

In this study, neutronic analysis of the HYLIFE-II reactor was investigated by inserting fuel rods containing UC or mixed ThC–UC into reflector zone partially. Four different coolants, namely, flibe, helium, natural lithium, and light water were considered in the fissile fuel breeding zone for comparison. Neutron transport calculations per incident (D,T) fusion neutron were performed by using the code Scale 4.3 under resonance-effect and resonance-free conditions. Numerical results pointed out that replacing the reflector zone by fissile fuel breeding zone even with a thickness of 14 cm improved the neutronic performance remarkably with respect to energy amplification and fissile fuel breeding.     

Investigation on the Neutronic Performance of a Fusion Reactor Using Flibe with Heavy Metal Fluorides

Using liquid wall between the plasma and solid first wall in a fusion reactor allows to use high neutron wall loads and could eliminate frequent replacement of the first wall structure during reactor’s lifetime. Liquid wall should have a certain effective or optimum thickness to extend solid first wall lifetime to reactor’s lifetime and supply sufficient tritium for deuterium–tritium (DT) fusion driver. This study presents the effect of thickness of flowing liquid wall containing 90 mol % Flibe+10 mol % UF₄ or ThF₄ on the neutronic performance of a magnetic fusion reactor design called APEX. Neutron transport calculations were carried out with the aid of code Scale4.3. Numerical results brought out that optimum liquid wall thickness of ∼38 cm was found for the blankets using Flibe+10% UF₄ whereas, 56 cm for that with Flibe+10% ThF₄. Significant amount of high quality fissile fuel was produced by using heavy metal salt.     

Rejuvenation of spent fuel in the force-free helical reactor (FFHR)

This study considers spent fuel rejuvenation potential of the force-free helical reactor (FFHR), which is a relevant heliotron-type D-T fusion reactor. For this purpose, three different spent fuels were selected, i.e.: (1) CANDU, (2) PWR-UO₂, and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the FFHR. The FZ was cooled with high-pressured helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years by 75% plant factor (η) under a first-wall neutron load (P) of 1.5 MW/m². The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δt) of 1/12 year (one month), by using the interface program (code). In all investigated spent fuel cases, the tritium self-sufficiency is maintained for DT fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU reactor after a regeneration time of ∼6 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO₂ and -MOX spent fuels after 40 and 34 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during spent fuel rejuvenation. Consequently, the blanket has high neutronic performance for the rejuvenation of the spent fuels.     

Assessment of criticality and burn up behavior of candu reactors with nuclear waste trans uranium fuel

Large quantities of nuclear waste plutonium and minor actinides (MAs) have been accumulated in the civilian light water reactors (LWRs) and CANDU reactors. These trans uranium (TRU) elements are all fissionable, and thus can be considered as fissile fuel materials in form of mixed fuel with thorium or nat-uranium in the latter. CANDU fuel compacts made of tristructural-isotropic (TRISO) type pellets would withstand very high burn ups without fuel change.As carbide fuels allow higher fissile material density than oxide fuels, following fuel compositions have been selected for investigations: ① 90% nat-UC + 10% TRUC, ② 70% nat-UC + 30% TRUC and ③ 50% nat-UC + 50% TRUC. Higher TRUC charge leads to longer power plant operation periods without fuel change. The behavior of the criticality k_∞ and the burn up values of the reactor have been pursued by full power operation for > ∼12 years. For these selected fuel compositions, the reactor criticality starts by k_∞ = 1.4443, 1.4872 and 1.5238, where corresponding reactor operation times and burn up values have been calculated as 2.8 years, 8 years and 12.5 years, and 62, 430 MW.D/MT, 176,000 and 280,000 MW.D/MT, with fuel consumption rates of ∼16, 5.68 and 3.57 g/MW.D respectively. These high burn ups would reduce the nuclear waste mass per unit energy output drastically. The study has show clearly that TRU in form of TRISO fuel pellets will provide sufficient criticality as well as reasonable burn up for CANDU reactors in order to justify their consideration as alternative fuel.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

Local power peaking factor estimation in nuclear fuel by artificial neural networks

This paper presents the training of an artificial neural network (ANN) to accurately predict, in very short time, a physical parameter used in nuclear fuel reactor optimization: the local power peaking factor (LPPF) in a typical boiling water reactor (BWR) fuel lattice. The ANN training patterns are distribution of fissile and burnable poison materials in the fuel lattice and their associated LPPF. These data were obtained by modeling the fuel lattices with a neutronic simulator: the HELIOS transport code. The combination of the pin U²³⁵ enrichment and the Gd₂O₃ (gadolinia) concentration, inside the 10 × 10 fuel lattice array, was encoded by three different methods. However, the only encoding method that was able to give a good prediction of the LPPF was the method which added the U²³⁵ enrichment and the gadolinia concentration. The results show that the relative error in the estimation of the LPPF, obtained by the trained ANN, ranged from 0.022% to 0.045%, with respect to the HELIOS results.     

Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium–tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li₂BeF₄ as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF₄ or UF₄) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S ₈–P ₃ approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF₄ or ThF₄ and 4% UF₄. In addition, higher energy multiplication values were found for the salt with UF₄ compared to that with ThF₄. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.     

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

A comparative neutronic study of the standard HEU core and various potential LEU alternatives for a typical MNSR system

A comparative study has been performed for neutronic analysis of highly enriched in uranium (HEU) and potential low enriched in uranium (LEU) cores for the Pakistan Research Reactor-2 (PARR-2) taken as a typical miniature neutron source reactor (MNSR) system. The group constant generation has been carried out using transport theory code WIMS-D4 and a detailed five-group RZ-model has been used in the CITATION code for multigroup diffusion theory analysis. The neutronic analysis of the 90% HEU reference and potential LEU alternative: UO₂, U₃Si₂ and U9Mo, cores has been carried out yielding 11%, 20.7% and 14.25% enrichments with corresponding values of excess reactivity: 4.33, 4.30 and 4.07 mk. These results have been found in good agreement with recently reported Monte Carlo-based transport theory calculations. The diffusion theory-based calculated values of thermal flux profiles for axial as well as for radial directions have been found to agree well with the corresponding experimental measurements. The UO₂-based LEU core has been found having flux spectrum closest to the reference core while U9Mo core has significantly harder flux spectrum at irradiation site.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the ⁹⁹TC and ¹²⁹I and ¹³⁵Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of ⁶Li in lithium (Li) and the theoretical density (TD) of Li₂O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ≥ 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.     

Neutronic analysis for core conversion (HEU–LEU) of Pakistan research reactor-2 (PARR-2)

Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl₄–Al). For core conversion, UO₂ fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO₂ of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel.     

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters in (D–T) driven hybrid reactor system

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters are investigated for different coolants under various first wall loads (P_w=2, 5, 7, 8, 9, and 10 MW m⁻²) in (D, T) (deuterium and tritium) driven and fueled with UO₂ hybrid reactors. Plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. The fissile fuel zone is considered to be cooled with four different coolants with various volume fractions, the volumetric ratio of coolant-to-fuel [(Vm/V_f) = 1:2, 1:1, and 2:1], gas (He, CO₂), flibe (Li₂BeF₄), natural lithium (Li), and eutectic lithium (Li₁₇Pb₈₃). Calculation in the fuel rods and the behavior of the fissile fuel have been observed during 4 years for discrete time intervals of Δt=15 days and by a plant factor (PF) of 75%. As a result of the calculation, cumulative fissile fuel enrichment (CFFE) value indicating rejuvenation performance has increased by increasing P_w for all coolants and . Although CFFE and neutronic performance parameter values increase to the higher values by increasing P_w, the maximum temperature in the centerline of the fuel roads has exceeded the melting point (T_m>2830°C) of the fuel material during the operation periods. However, the best CFFE (11.154%) is obtained in gas coolant blanket for =1:2 (29.462% coolant, 58.924% fuel, 11.614% clad), under 10 MW m⁻² first wall load, followed by flibe with CFFE=11.081% for =2:1 (62.557% coolant, 31.278% fuel, 6.165% clad), under 7 MW m⁻², and flibe with CFFE=9.995% for =1:1 (45.515% coolant, 45.515% fuel, 8.971% clad), under 7 MW m⁻² during operation period without reaching the melting point of the fuel material. While maximum CFFE value has been obtained in fuel rod row#10 in gas, natural lithium, and eutectic lithium coolant blankets, it has been obtained in fuel rod row#1 in flibe coolant blanket for all and P_w. At the same condition, the best neutronic performance parameter values, tritium breeding ratio (TBR)= 1.4454, energy multiplication factor (M)= 9.2018, and neutron leakage (L)= 0.0872, have been obtained in eutectic lithium coolant blankets for the =1:2, followed by gas, natural lithium, and flibe coolant blankets. The isotopic percentage of ²⁴⁰Pu is higher than 5% in all blankets for P_w 7 MW m⁻², so that plutonium component in all blankets can be never reach a nuclear weapon grade quality during the operation period.