Calculation of fuel burn up and radioactive inventory in a CANDU reactor using WIMS-D4 code

A study of fuel burn up and radioactive inventory for the CANDU reactor is carried out to validate the computer code WIMS-D4 for cluster geometry. The infinite and effective multiplication factors are calculated as a function of burn up and compared with the available results. A good agreement is observed among the present calculations and the previous published results. The code is then used to calculate the ²³⁵U and ²³⁸U consumed and the ²³⁹Pu produced in the fuel bundle. The inventories and the corresponding activities of some important fission products are also calculated.     

Criticality safety of the ET-RR-1 new spent fuel storage pool

A new ET-RR-1 spent fuel storage pool is now under construction on the reactor site at Inshass. In addition, the pool is designed to accommodate spent fuel of MTR type as well. Criticality safety of this pool for the different fuel types has been evaluated as a function of U²³⁵ loading. The effect of fuel element separation (rows and columns) on the eigenvalue has been studied. As a conservative assumption, the pool is assumed to be filled with fresh fuel. The eigenvalue considering a realistic degree of fuel burn-up was determined in order to determine the safety margin. The calculations have been carried out using the code packages of the National Center for Nuclear Safety and Radiation Control.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Numerical studies of nuclear traveling waves in a supercritical water cooled fast reactor

An axial fuel shuffling strategy is proposed based on the mechanism of the nuclear fission traveling wave and implemented numerically in the calculation for a supercritical water cooled fast reactor (SCWFR). The ERANOS code is adopted to perform the neutronics and burn-up calculations, and the calculation scheme for axial fuel shuffling and coolant density coupling are set up. The parametric studies of a typical PWR with Th-U and U-Pu (²³⁵U instead of ²³⁹Pu) conversions by burn-up and k_eff calculations indicate that the breeding effects only exist in configurations with very low water content and the conversion or breeding becomes worse as the initial enrichment is increasing. The shuffling calculations for the 1-D SCWFR model described in this paper brought about some interesting results for a certain range of water content. The results indicate that the non-enriched fresh fuel is not possible for both Th-U and U-Pu conversions. As could be expected due to the η-values of the main fissile isotopes ²³³U and (²³⁵U, ²³⁹Pu), respectively, the Th-U conversion needs a lower enrichment, and results in a slightly higher burn-up than the U-Pu conversion. The asymptotic power density distribution of the Th-U conversion is broader and lower than that of the U-Pu conversion. By reducing the water volume fraction, an increased burn-up can be achieved with correspondingly reduced fuel shuffling speed and reduced initial enrichment. Furthermore, the steady state calculations for the asymptotic state show that the Th-U conversion is superior to the U-Pu one concerning SCWFR safety aspects, where the absolute value of the Doppler constant is larger and the coolant feedback is negative for the Th-U conversion, while the coolant feedback is positive for the U-Pu one.     

The effect of high pressure test loop (HPTL) on the Reactivity of ET-RR-2 reactor

The second Egyptian Research Reactor ET-RR-2 is a multipurpose research reactor. It is an open pool type, with nominal power of 22 MW water-cooled. The reactor pool is designed to accommodate two fuel test loops mainly 500 and 20 KW loop in the reactor reflector to enable performing experiments on the behavior of fuel rods for nuclear reactors under their operating conditions. For that, inserted high-pressure test loop (HPTL) loaded with suggested CANDU type fuel element in the reactor core is important to achieve the above reason. From the neutronic safety point of view, it is necessary to study the mutual neutronic and reactivity effect between the reactor core and HPTL. This paper aimed at the study of the temperature coefficients of fuel and moderator of the CANDU type fuel element at different ²³⁵U enrichments, and the effect of HPTL on the reactor core reactivity. The effect of flooding the contact second shut down system (SSS) chamber with water and gadolinium nitrate on the reactor core reactivity in the presence of HPTL. All analysis was performed with the WIMSD4 and DIXY2 codes. This study shows that, an unacceptable change of reactor core reactivity was found due to the presence of the HPTL and the maximum inserted reactivity does not exceed 527 pcm at high possible ²³⁵U enrichment (10%).     

Plutonium credit in the low enriched uranium fuel of PARR-1

The amount of plutonium (Pu) isotopes and the resultant savings of ²³⁵U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations.The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 × 10⁻³Δk/k. Amount of ²³⁵U equivalent to this value of reactivity was found to be 15.58 ± 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box.     

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.     

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd₂O₃. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

Comparative Studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 Data Libraries on the Monte Carlo Continuous Energy Modeling of the Gas Turbine-Modular Helium Reactor Operating with Thorium Fuels

One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile ²³²Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of ²³⁹Pu, ²³³U and ²³⁵U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B.     

Burnup characteristics and fuel cycle economics of mixed uranium–thorium fuel in a simplified small pebble bed reactor

The limitation of natural uranium resources and the improvement of economic values of nuclear reactors are important issues to be solved in the future development of these reactors. In our previous study, we presented an innovative design for simplifying a pebble bed reactor, and the optimization of this design showed that burnup values could be increased and natural uranium uses could be reduced. The purposes of the current study were to design a simplified pebble bed reactor by removing the unloading device from the reactor system and to further optimize the burnup characteristics of this reactor with a peu à peu fuel-loading scheme by introducing thorium in the fuel configuration as a fertile material. Another goal was to optimize the fuel composition so that the system could achieve even better burnup characteristics and use scarce uranium resources more efficiently. Using a specially developed computer code, we analyzed and optimized the performance of a 110-MWt simplified pebble bed reactor using a peu à peu fuel-loading scheme. An optimized design using 30% of fertile thorium mixed with uranium fuel with 15% ²³⁵U enrichment and a 7% packing fraction calculated to achieve a high burnup of 140 GWD/T for more than 21 years' operation time that could save 13 to 33% of natural uranium use compared with the savings noted in our previous study. Neutronic, burnup and fuel economic analysis for this optimized design are discussed in this study.     

Fuel lattice design in a boiling water reactor using an ant-colony-based system

This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U²³⁵ enrichment and Gd₂O₃ concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 × 10 BWR fuel lattice with an initial average U²³⁵ enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U²³⁵ enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U²³⁵ enrichment; whereas, the k-infinity was inside the ±100 pcm’s range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a powerful tool to tackle this step of the fuel design.     

Determination of enrichment of recycle uranium fuels for different burnup values

Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000–50,000 MWd/tonU.     

Calculation of the time for establishing radiation equivalence of high-level wastes

The potential harm of long-lived radionuclides estimated as the product of the activity of a radionuclide and the dose coefficient is examined. The potential harm of actinides in high-level wastes is calculated taking account of the harm due to their decay products. At the same time, an analogous calculation is performed for the uranium isotopes ²³⁸U, ²³⁵U, and ²³⁴U consumed in the reactor. The value obtained, which depends on the time, can be regarded as the averted harm. The time for establishing radiation equivalence between the high-level wastes and the consumed uranium is determined as the time in which the potential harm from actinides becomes equal to the averted harm. It depends on the holding time of the spent fuel before radiochemical reprocessing. For a 5 yr holding period, it is 49000 yr.__________Translated from Atomnaya Énergiya, Vol. 98, No. 2, pp. 123–129, February, 2005.     

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.     

Development and verification of fuel burn-up calculation model in a reduced reactor geometry

A fuel burn-up model in a reduced reactor geometry (2-D) is successfully developed and implemented in the Batan in-core fuel management code, Batan-FUEL. Considering the bank mode operation of the control rods, several interpolation functions are investigated which best approximate the 3-D fuel assembly radial power distributions across the core as function of insertion depth of the control rods. Concerning the applicability of the interpolation functions, it can be concluded that the optimal coefficients of the interpolation functions are not very sensitive to the core configuration and core or fuel composition in RSG GAS (MPR-30) reactor. Consequently, once the optimal interpolation function and its coefficients are derived then they can be used for 2-D routine operational in-core fuel management without repeating the expensive 3-D neutron diffusion calculations. At the selected fuel elements (at H-9 and G-6 core grid positions), the discrepancy of the FECFs (fuel element channel power peaking factors) between the 2-D and 3-D models are within the range of 3.637 × 10⁻⁴, 3.241 × 10⁻⁴ and 7.556 × 10⁻⁴ for the oxide, silicide cores with 250 g ²³⁵U/FE and the silicide core with 300 g ²³⁵U/FE, respectively.     

Estimation of Neutronic Performance in a Hybrid Reactor with Regression Analysis

This study presents regression analysis method used for prediction and investigation of neutronic performance in a hybrid reactor using UO₂ fuel and Flibe (Li₂BeF₄) coolant. The ²³⁵U fraction is increased gradually from 0 to 4% stepped by 1% and the ⁶Li fraction within the Flibe coolant is enriched gradually to 30, 60 and 90% from 7.5%. Relations between ²³⁵U fuel fraction and lithium (⁶Li) enrichment are investigated for the estimation of neutronic performance as the tritium breeding ratio (TBR), energy multiplication factor (M), total fission rate (Σ_f), ²³⁸U (n,γ) reaction and fissile fuel breeding (FFB) in the hybrid reactor. Regression analysis by results obtained by using the code (XSDRNPM/SCALE5) for TBR, M, Σ_f, ²³⁸U (n,γ) and FFB are performed. The results of the regression analysis and the values obtained by using the code (XSDRNPM/SCALE5) are compared with respect to the TBR, M, Σ_f, ²³⁸U (n,γ) and FFB of the reactor. The values calculated from the obtained formulations with regression analysis are found to be in good agreement with results obtained by using the code (XSDRNPM/SCALE5). It is observed that the derived equations from regression analysis could provide an accurate computation of the neutronic performances so that these equations could use for the prediction of TBR, M, Σ_f, ²³⁸U (n,γ) and FFB. In addition, correlation matrix is calculated to determine the degree of relationship between variables as TBR, M, Σ_f, ²³⁸U (n,γ) and FFB.     

Fission MOLY production at PARR-1 using LEU plate type target

Neutronic analyses and depletion calculations have been performed for the production of ⁹⁹Mo at PARR-1. Analysis has been performed with 20% enriched ²³⁵U target bearing plate type aluminized fuel (U-Al). Target (target holder and fuel plates) design contains three fuel plates and two aluminum dummy plates. Neutronic calculations were carried out for core at the beginning of equilibrium cycle of Pakistan Research Reactor-1 (PARR-1). Target analysis was performed by irradiating it at a location of maximum thermal flux available in the core. For this purpose, irradiation was performed at five different axial planes of the central water box facility. Thermal neutron flux profiles were also studied at different axial positions in available irradiation locations. Computer code WIMSD/4, a transport theory lattice code was employed for the generation of 10 group microscopic cross-sections. Diffusion theory code CITATION was utilized for three-dimensional modeling of the core. It was observed that from reactivity point of view, insertion or removal of target from the core will not affect the safety of reactor. Maximum heat flux in the target would be 102.68 W/cm² which is below the point of onset of nucleate boiling. However, forced flow is required to avoid initiation of nucleate boiling. The computer code ORIGEN2 was employed for depletion calculations. Analysis was performed for 100 Ci activity of ⁹⁹Mo. After 100 days decay, waste activity will be less than 1 Ci and it will not pose any problem for handling radioactive waste.     

Lowering the enrichment of the Syrian miniature neutron source reactor

An investigation of lowering the fuel enrichment of MNSR was realized. A 3-D neutronic model was developed for the analysis of the reactor. It was found that lower number of fuel elements is needed when UO₂ is used with 5.45 g of ²³⁵U content in each fuel rod. Sensitivity of the reactor to the purity of the beryllium reflector proved to be an important factor in determining the reactor neutronics as well as the weight of loaded fuel in the core. Inherent safety features of low excess reactivity and shutdown margins are preserved and enhanced. Average thermal fluxes over different zones of the core are kept very much unchanged. ©