Enriched UO2 Research Articles

Neutronic analysis of new generation TVS-2M fuel assemblies in VVER-1000 reactor type

The Russian fuel assembly TVS-2M is a newly proposed fuel configuration that contains gadolinium-oxide at different concentrations mixed with enriched UO2 at different U-235 enrichment values. The purpose of this study is to investigate the neutronic behavior of the Bushehr VVER-1000 reactor core by replacing the current standard Fuel Assemblies (FAs) called TVS with the new generation of TVS-2M FAs. In such manner, the reactor core was simulated under both modes of TVS-2M and TVS FAs arrangement using the WIMS, CITATION, MCNPX, and SuperMC codes in equal terms. Neutronic parameters of the reactor core such as effective multiplication factor, excess reactivity, neutron flux distribution, power peaking factor, and control rod worth are calculated and discussed within a comparative framework of both configurations of FAs; TVS and TVS-2M.

Potential of digital autoradiography for characterization of uranium materials

This study evaluates the capabilities of a commercial digital autoradiography set-up in characterizing the spatial distribution of radioactivity in U-bearing materials. Two different imaging plates (IPs) from various manufacturers, as well as three read-out granularities (25 µm, 50 µm, and 100 µm) of a high-resolution IP scanner, were tested. A custom software, developed using Python, was used for processing of the image raw data obtained from the IP scanner. The software extracted 25 intensity profiles from each autoradiographic image, allowing for a detailed analysis of the data. By applying various mathematical fitting functions, important metrics such as edge width, amplitude of the flat top, and central core width of the intensity profiles were determined. This provided valuable insights into the interaction between the U samples and IPs, particularly for samples of different geometry and thickness. Comparisons were made between samples embedded in epoxy resin or modelling clay and non-embedded U specimens. The results highlighted the superior performance of epoxy embedding in the characterization of U-bearing materials. The variation in signal intensity between runs for individual samples reached up to 20 %, showing consistent behavior across all samples in each measurement. This suggests that the inconsistency in signal response during repetitive analysis is likely due to the IP scanner rather than the IP itself. By using the analytical method of global spatial autocorrelation (Moran’s I), we were able to effectively detect enrichment inhomogeneity within low enriched UO2 pellets. The Moran’s I value quantified the levels of inhomogeneity, providing a numerical measure that was not visually apparent from the digital autoradiographic image.

Investigation of the Thermal Effects of MgO and ZnO Nanoparticles in a Pressurized Water Reactor Coolant with Computational Fluid Dynamics Model

When nanofluids are used as reactor coolants, they provide more effective heat transfer with their increase in thermal conduction properties. This plays an important role in energy production by increasing the efficiency of nuclear reactors. The present study delves into the thermal-hydraulic ramifications of utilizing nanofluids as coolants in the VVER-1000 nuclear reactor. Specifically, the thermal-hydraulic characteristics, encompassing parameters such as coolant temperature and departure from nucleate boiling ratio, were scrutinized in light of the incorporation of magnesium oxide (MgO) and zinc oxide nanoparticles. While performing these analyses, not only uranium but also thorium was used in the core as reactor fuel. Considering the emergence of thorium as a potential fuel material in nuclear technology, its inclusion in the fuel composition contributed to the originality of the research. With the addition of 0.2% MgO nanoparticles to a VVER nuclear reactor using 5% thorium dioxide (ThO2) fuel, the coolant temperature as a result of the channel flow was determined as 617.4 K (while 613.7 K for the light water). When employing thorium fuel (with an equivalent nanoparticle concentration), the maximum temperature exhibited an approximate increase of 3 deg compared to uranium fuel. With the addition of 0.2% MgO nanoparticles, the enthalpy value at the end of the channel was 1303.6 kJ/kg when using 5% ThO2 fuel, while the enthalpy value was determined as 1295 kJ/kg in 3.7% enriched UO2 fuel. As one of the most important results of the analysis, it was observed that the temperature value of the coolant increased when nanoparticles were used.

Implementation of thorium-based fuel for Indonesia Micro Reactor (IMR)

Indonesia has committed to reducing greenhouse gas emissions in the framework of net zero emissions, and nuclear energy will be a part of the energy mix. Indonesia has been developing a 10 MWt micro reactor of high temperature gas-cooled reactor (HTGR) with pebble fuel containing 17 % enriched UO2 kernel with OTTO cycle scheme. For the government’s interest to use thorium, this paper assesses the possibility of replacing uranium with thorium. The assessment is performed by calculating the effective neutron multiplication factor (keff). Neutrons captured by 232Th lead to the formation of two intermediate isotopes, namely 233Th and 233Pa, the latter being a significant neutron absorption cross-section. Neutronics analysis is performed using parameters of thorium mass fraction, core temperature, and active core height. The thorium mass fractions varies from 0 % to 90 % Th. The core temperatures are in the range of the normal operating temperature at 900 K, assumed superheat temperature at 1200 K, and shutdown temperature at 300 K. The active core height varies at 197, 180, 150, 130, and 125 cm. This paper aims to determine the keff of the IMR at maximum thorium/uranium mass fraction composition, maximum active core height, core temperature, and fuel burn-up. The MCNP code is used for analysis. The results show that keff decreases with increasing thorium mass fraction and core temperature, and decreasing active core height. Calculations with thorium mass fraction of 15 %, core active height of 197 cm in 20 steps show that the reactor may be continuously operated for more than 920 days without refueling. The analysis results show that IMR design has negative temperature reactivity, inherent safety characteristics, and the ability to use thorium-based fuels.

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

This paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO2. The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB2 Integral Fuel Burnable Absorber (IFBA) and B4C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd2O3-doped UO2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

Uncertainty analysis of the China experimental fast reactor start-up tests based on refined calculation results

In this paper, the numerical calculations of China Experimental Fast Reactor (CEFR) start-up tests were performed by using SARAX code with the improved core model. The results of criticality, control rod worth, and reactivity effect were compared against the experimental data. The numerical results showed the SARAX code could be reliably used for CEFR. Moreover, the sensitivity of the keff value due to the change of input parameters was analyzed. At the same time, the uncertainty quantification of calculated keff value and control rod worth and detailed contribution to the total uncertainty were analyzed. For the reactor loaded with highly enriched UO2 fuel, the uncertainty of keff value could reach more than 2%. Besides, the capture reaction of 235U contributed the most with a value of 1.91%. The uncertainty analysis of the control rod worth showed that the total uncertainty depended on the control rod layout rather than the composition.

Investigating the possibility of extending the BWR cycle length for 15 years of operation by mixing highly enriched UO2 fuel with burnable absorbers

This paper investigates the possibility to extend the cycle length of boiling water reactor bundles to 15 years of operation with three different burnable poisons; gadolinium, erbium, and boron carbide. This can be carried out by mixing highly enriched UO2 fuel (15–19.9% U-235) with high concentrations of Gadolinium oxide (3–14% Gd2O3) or Erbium oxide (2–4% Er2O3).The boron carbide B4C was modeled as (Al2O3-B4C) rods in the bundle guide tubes. MCNPX code 2.7 was used to evaluate infinite multiplication factor (K-inf), power distribution, peaking factor, void reactivity coefficient, fuel cycle length, depletion of U-235, and fissile inventory ratio for the three designs at 40% void. The MCNPX simulation showed that introducing gadolinium rods at the bundle periphery has the advantage of lowering reactivity swing throughout the exposure range. The uniform distribution of erbium in all fuel rods contributed to the flattening of peaking factor at all the burnup stages. For the B4C design, the author found that the assembly with B4C–Al performs best in terms of reactivity flattening when five of the B4C–AL2O3 rods are positioned in the central region of the assembly. Furthermore, the fuel temperature coefficient is more negative for gadolinium design at all burnup stages. On the other hand, the boron model delivers the lowest control rod worth. Finally, the moderator temperature coefficient is more negative for erbium and WABA designs due to the enhanced thermal neutrons capture by the effect of the strategic arrangement of WABA rods and the uniform distribution of erbium.

Developing a Reactivity-Equivalent Physical Transformation to Simulate an Axially Heterogeneous Boiling Water Reactor

Hitachi is advancing their designs for a conceptual reactor called the resource-renewable boiling water reactor (RBWR), a concept reactor similar to the advanced boiling water reactor with a harder neutron spectrum. This design aims to minimise construction costs and waste production as well as to utilise separated plutonium and minor actinide fuel. However, the axial heterogeneity of the design poses calculation difficulties. The aim of this work is to use a known method, reactivity-equivalent physical transformation (RPT), for calculating fuel with double heterogeneity and apply it to a BWR-type fuel pin. This could reduce the calculation time needed for optimisation of the design of the RBWR. The objective of the study is to use SCALE 6.2 to produce an equivalent axial pin model by comparison with the burnup and neutron spectra of a radial model of the fuel. This model can then be used for 2D burnup calculations, and in future work will be used for the generation of two-group and multigroup cross-sections for further deterministic calculations as part of a two-step approach for analysis of the RBWR. The RPT method has been extensively tested on spherical fuel, and SCALE is a standard industry code. The initial radial model is a hexagonal assembly with 20% enriched UO2 fuel in a zircaloy cladding, surrounded by light water moderator. The derived axial model has a water density distribution taken from Hitachi’s RBWR designs. Criticality over 70 GWd/tU burnup is estimated using the model. The application of the RPT to the BWR pin was shown to be possible, but to have limitations with the introduction of additional radial complexity. For a single pin, excellent agreement between the radial and axial models could be found across a range of water densities, but in the case of an assembly level calculation distinct equivalence models were required for each water density. In addition, the produced RPT model is validated using SCALE’s 3D Monte Carlo module, KENO.

Operational characteristics of LWR core fuelled with thorium-based fuel

The characteristics of a Pressurised Water Reactor (PWR), considered representative Light Water reactor (LWR), operating on thorium-based fuel has been analysed. An equilibrium cycle of PWR consisting of 151 fuel assemblies and having three-batch fuel management has been considered for the analysis. The two types of thorium-based fuels resulting from mixing LEU and plutonium with ThO2 and representative UO2 have been taken for the core loading and fuel cycle study. The reactivity coefficients, kinetic parameters over the equilibrium cycle have similar trend for three fuel types considered. The control inventory with the natural boron in B4C and in dissolved boric acid is sufficient for the safe operational reactivity control and shutdown in all three alternatives. We find that enriched UO2 based fuel in conventional PWRs can be successfully replaced with ThO2 with Plutonium or LEU as fissile feed. Such approach is useful in initiating thorium as nuclear fuel which will support other diverse approaches requiring design and dimension changes for the utilization of thorium.

A sensitivity analysis to predict the neutronics behavior of samples irradiated in the VTR rabbit system

A low-order neutronics model is developed to carry out hundreds of simulations efficiently and investigate the neutronics behavior of samples being irradiated in a test reactor setting under different geometrical constraints. The low-order model allowed for simulations that yield the expected neutronics behavior of any irradiated sample in any environment and allows for the calculation of highly accurate spatially averaged statistics and idealized spatial distributions in the neutron flux. Several benchmarks are performed to evaluate the performance and limitations of the low-order model revealing many important findings. The low-order model predicted the LHGR in the EBR-II driver fuel to within 2.34% by only simulating the fuel rod by itself, which served as a validation for the model. Sensitivity studies investigated 3% enriched UO2 and U-10Zr being irradiated in the Versatile Test Reactor rabbit system. The analyses investigated a range of combinations of 15 radii and 5 heights for each sample in the rabbit system. Similar data sets are also provided for irradiations in the Advanced Test Reactor’s B-10 irradiation position, which is a thermal neutron spectrum environment. Generalized fits and fit coefficients are obtained for sample heating, reaction rate densities, and local multiplication rate characteristics, allowing the predictions of the neutronics behavior of the samples based on their geometrical constraints. The analyses and fits laid the groundwork for developing a user-end Multiphysics analysis framework to assist and accelerate irradiation experiment design and optimization.

Wetting-induced volumetric collapse of UO2 powder beds and the consequence on transient nuclear criticality excursions

Mathematical and computational models are proposed to simulate wetting-induced volumetric collapse of fissile powder beds. Slumping, nuclear thermal hydraulics, radiolytic gas, and steam production models are coupled with point neutron kinetics to investigate transient nuclear criticality excursions in two 5-wt% enriched UO2 fissile powder beds with varying levels of wetting-induced volumetric collapse. The two beds are distinguished by their mean powder particle size of 30 μm and 100 μm. For the UO2 powder beds modelled, the re-distribution of UO2 powder and moderator due to slumping introduced a negative reactivity into the system. This increased the amount of time taken for a delayed critical state to be reached once infiltration began, and also reduced the total fission energy generated over the course of the simulated transient. The total fission energy generated ranged from 42 MJ to 48 MJ 100 seconds after the initial nuclear criticality excursion was observed for the 30μm sized UO2 powder bed. The fission energy of the larger sized powder bed (100 μm), varied from 42 MJ to 57 MJ. Larger discrepancies between the slumped and un-slumped initial peak power are predicted. Peak powers varied from 29.2 MW to 106 MW for the smaller-sized powder particles, whereas for larger particles, the peak powers varied from 255 MW to 501 MW.

Design and validation of a gamma-ray scanning system for measuring irradiated nuclear fuel

A prototype gamma-ray scanning system was designed to perform post-irradiation examination measurements of nuclear fuel. The system is composed of a coaxial high-purity germanium (HPGe) detector, a collimator, and mechanical positioning stages that translate the fuel sample across the front of the collimator. This paper documents the design process, modeling and the laboratory characterization performed to validate the system. The modeling of the system was done using the Monte Carlo N-Particle code with vendor-supplied design specifications, dimensional measurements, and X-ray radiographs of the HPGe detector. The characterization included measurements of calibrated radiation sources in different geometries, and calculations for absolute efficiency and dead time. Benchmark measurements were performed by scanning an irradiated fuel rodlet containing ten pellets of 4.9% enriched UO2 in zirconium alloy. These results were compared to simulated spectra to further help characterize the detector model. One goal of the system is to determine the number of fissions per gram of UO2 in the fuel; comparison of the results from this prototype system's assay of the irradiated fuel rodlet are in agreement with previously estimated results.

Mathematical and computational models for simulating transient nuclear criticality excursions within wetted fissile powder systems

This paper describes a novel methodology for the analysis of nuclear criticality excursions in fissile powder beds under wetting conditions. These potentially hazardous powder, slurry and sludge systems may be found in nuclear fuel manufacturing and fabrication facilities. A point neutron kinetics model was coupled with water infiltration, thermal–hydraulics and radiolysis models through the use of reactivity feedbacks. Good agreement in the water infiltration rate was found when comparing the water infiltration model used in this paper to experiments conducted by the French Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA). A case study was proposed whereby a sheet of fine water droplets from a sprinkler system came into contact with an open-topped bed of low enriched UO2 powder. Simulations indicate that the mean powder particle size had a strong effect on the time required for the water to percolate through the powder bed. Powder particle size was also predicted to have a moderate effect on the initial fission power spike. The fission energy released over the first 300 s of the nuclear criticality transient ranged from 65.28 MJ to 97.98 MJ depending on mean powder particle size. This is similar in magnitude to other simulated nuclear criticality excursions in powder beds. The model predicts that the initial fission power spike would be limited by the production of radiolytic gas and to a lesser extent the effects of Doppler broadening and thermal expansion. As expected, boiling and the associated steam production, was found to be an important phenomenon in the reduction of the fission rate through the negative void reactivity effect of the steam.

Reactor core design of UPR-s: A nuclear reactor for silence thermoelectric system NUSTER

The heat pipe reactor is featured with the passive heat discharge, anti-single point failure and silent operation. Therefore, it could be applied for the power system of underwater vehicles. In this paper, the concept of Unattended Uranium-fueled Portable Reactor (UPR) for the NUclear Silence ThermoElectric Reactor (NUSTER) power system is introduced and the reactor core with a thermal power of 1 MW is neutronics designed. In the core, highly enriched UO2 fuel, sodium heat pipe, molybdenum alloy matrix, beryllium oxide and beryllium reflector are adopted and the reactivity is controlled by four sliding reflectors and four safety rods. In design process, the NECP-MCX code is used to carry out the neutronics calculation. At the same time, the heating unit, core loading, control system and power distribution of the core are described. Furthermore, criticality safety, burnup characteristics, power distribution of the core are specifically analyzed. The results show that the proposed core design meets the requirements of criticality safety and operating life.

Modelling of reusable target materials for the production of fission produced 99Mo using MCNP6.2 and CINDER90

Current fission-based methods of 99Mo production require single use uranium targets which leads to spent uranium waste. This waste could be reduced if a target is developed that does not require dissolution so that it can be reused for multiple production runs. MCNP6.2 was used to model reusable targets of 20% and 1% enrichment for activity produced, target efficiency and burnup. The 1% enriched target was found to be much more efficient but had a lower activity produced compared to the 20% enriched target. The ideal target design for 99Mo production that optimises efficiency and reusability and reduces the self-shielding effect of UO2 was found to be a target that is made from 1% enriched UO2 with density as high as allowable for sufficient yields, efficient 99Mo extraction and having an irradiation time of 5 days, with the target able to be re-irradiated and re-processed 2–4 times.

Reactor design of UPR-1: A MW-level reactor for unmanned underwater vehicle propulsion using low-enriched uranium

Unmanned underwater vehicles (UUVs) are being used for many civil and military applications. Power sources with longer life are required for deployment in UUVs. Unlike conventional power sources, nuclear reactor with longer life can be considered for such application. In the present paper, a MW-level Lead-cooled reactor is proposed for UUV power source. Low enriched UO2 core with 10 EFPY life is considered for the present design. Various core configurations have been considered to obtain the final optimized design. This design satisfies the required criteria. Enough excess reactivity with adequate shutdown margin is ensured to run the system for its design life. The shielding performance has also been evaluated to show the feasibility of nuclear source in a compact unmanned facility.

Validation of deterministic code DRAGON5 for the fuel depletion analysis of a PWR pin-cell benchmark

The purpose of this paper is to investigate the DRAGON5 accuracy for studying a PWR pin-cell benchmark, using the ENDF/B-VIII.0 nuclear data library based on 172 energy groups, in DRAGLIB format. This paper focuses on the neutronic and burn-up analysis of the Uranium fuel using DRAGON5 transport code. The precision of the deterministic Code DRAGON5 was validated by comparing Results obtained from analysis of the UO2 pin cell high burn-up benchmark with previously published literature values. The results of a study showed good agreement. Maximum absolute difference in infinite multiplication factor between DRAGON5 and CASMO-4 is found to be 0.23% for UO2 pin cell benchmark. Furthermore, the isotopic inventories and activities have been calculated/evaluated versus the fuel burn-up (BU) time for the natural UO2 fuel, and the enriched UO2 fuel at 1.2%.

Use of americium as a burnable absorber for VVER-1200 reactor

The objective of this research is to the use of americium (AmO2) as a burnable absorber effectively instead of conventional gadolinium (Gd2O3) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO2 as a burnable absorber. From these studies, it is found that AmO2 can decrease the excess reactivity like Gd2O3 without changing the criticality life span and enrichment of 235U. A homogeneous mixture of the 0.20% AmO2+ 4.95% enriched UO2 fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO2+UO2 in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

Non-Destructive post-irradiation examination results of the first modern fueled experiments in TREAT

A series of experiments were conducted in the Transient Reactor Test (TREAT) Facility as part of commissioning for nuclear fuel safety research to be continued over the next several decades. TREAT was first constructed in the late 1950s to support research on nuclear fuel specimens under extreme nuclear-heated conditions. Following a decades-long hiatus, reactor operations were resumed at TREAT in 2017 to support the reemerging field of fuel-safety research. The first fueled tests were performed on fresh light-water reactor (LWR)-type sub length specimens (4.9% enriched UO2 in zirconium-alloy cladding) in inert gas capsules with the primary objective to demonstrate new irradiation capabilities, including calculation of energy-coupling factor, and to provide a baseline reference for interpretation of future experiments on accident tolerant fuel designs and high burnup standard fuels. In parallel, several new nondestructive post-irradiation experimental systems and processes have been developed and were successfully demonstrated using these samples. Five capsules were irradiated with energy injections and peak cladding temperatures of ~212 and ~1312 J/gUO2 and ~561–~2113 °C, respectively. These irradiations resulted in various degrees of fuel damage that were characterized using advanced non-destructive examinations. The post-irradiation examinations highlighted that, in the most-energetic transient, zirconium breakout, also referred to as ‘candling,’ was observed as gross cladding melting and relocation. Post-irradiation examinations also confirmed results inferred from in situ instrumentation during irradiation and are consistent with previous results from other studies.

3D Analysis of a multipurpose modular reactor for small energy production

The Fixed Bed Nuclear Reactor (FBNR) is a modular reactor for small energy needs (70-120 MWel). FBNR is pressurized water reactor, but uses GEN-IV reactors fuel. UO2 kernels of 500 μm diameter with 25 μm thick SiC cladding are embedded in graphite matrix form the spherical fuel elements with a diameter of 15 mm and covered by 300 μm thick Zircaloy-4 cladding. FBNR has high level safety features and operates without On-site Refueling. Approximately 1.62 million spherical fuel elements are suspended in a core height of 200 cm and inner and outer diameters of 32 cm and 224 cm by upwards flowing light water coolant with 280°C entry and 340°C exit temperatures. The unit lattice geometry is dodecahedron with fuel element in center and surrounded with light water coolant-moderator.In the present work, 9% enriched UO2 fuel is used. Neutronic analysis has been performed with the MCNPX-2.7.0 code. As, it is not practical to tackle millions of fuel elements separately, at first run unit fuel cell calculations are performed in spherical geometry for one single lattice consisting of separate fuel, cladding and volume-equivalent moderator regions. In the second run, the fuel cell is homogenized. Infinite cell calculations have been executed for variable moderator/fuel (H/235U) ratios in order to determine under- and over-moderated criticality values. The criticality values for the homogenized cell geometry are slightly lower and remains <5% range, that is, on the conservative side, which is acceptable for the purpose of this study. This allows us to homogenize the entire core for the full 3-D reactor with core and reflector regions. Reactor criticality increases with increasing H/235U ratios in the, under moderated region, where H/235U = 18 is selected for further investigations. Temporal variation of the reactor criticality is pursued for full reactor power of 210 MWth (~70 MWel net) with keff = 1.36 at the beginning of life down to keff = 1.02 over 1260 days with a plant factor of 1.0. The final fuel burn up was found as 57.3 GWd/ton.