Void Reactivity Coefficient Research Articles

Examination of the use of thorium-based fuel for burning minor actinides in European sodium cooled fast reactor

Abstract To reduce the potential and volume of radiotoxicity of radioactive wastes, minor actinides (MAs) elements recovered from discharged fuel are to be recycled. In this context, the main aim of this work is to examine the possibility of using thorium-based fuel for burning extra MAs in the fourth generation European Sodium cooled Fast Reactor (ESFR). For that MCNPX transport code was used to design a representative fuel assembly of this concept. The reference uranium-based fuel was uniformly replaced by thorium mixed with different fractions of MAs and burned for 2050 effective full power days. The influence of MAs on the evolution of reactivity and fuel transmutation is being investigated to ensure the performance of the fuel. Also, different neutronic and safety parameters such as neutron yield, neutron spectrum, neutron flux, effective delayed neutron fraction, Doppler constant, sodium void reactivity coefficient, were computed and their change with MAs content was investigated. Results indicate that recycling of MAs in ESFR using thorium-based fuel has a positive effect on reactivity and transmutation evolution but on other hand has a negative effect on safety parameters.

Burnup and neutronic parameter analysis of GAMA molten salt reactor (GAMA-MSR)

The GAMA molten salt reactor (GAMA-MSR) is a small modular MSR design developed by the Nuclear Reactor Research Team of the Department of Nuclear Engineering and Engineering Physics, Universitas Gadjah Mada. In the GAMA-MSR design, the reactor core is placed above the primary heat exchanger. The primary heat exchanger is partially filled with fuel in normal operation, allowing room to be filled with the fuel drained from the reactor. Thus, in addition to control rods, the design of GAMA-MSR provides unique reactivity control through the balance of fuel salt fuel from the reactor core to the primary heat exchanger vice versa by adjusting fuel pump capacity. This design allows to shutdown the reactor by switching off the fuel pump. The reactor is initially fueled with 7LiF-ThF4-UF4 with 75: 20.1: 4.9 mol composition. The uranium has 19.75 % U-235 mol fraction. The reactor criticality and nuclide composition in the reactor fuel are calculated using OpenMC code during the reactor operation lifetime, with periodic adjustments to the fuel composition to simulate the effects of the fissile and fertile additions to the fuel and the extraction of actinides and fission products from the fuel. The calculated reactor thermal power is 60 MWth in this paper. The reactor is operated until it achieves a burnup value of 34,000 MWd/tHM. At this burnup value, the U-235 inventory is consumed and the amount is reduced from 794 kg to 250 kg. The U-233, Pu-239 and Pu-241 are produced with the amount are 246 kg, 24 kg and 8 kg respectively at the end of reactor operation. The effects of temperature, fuel density change, and fission product neutron poisons are also calculated. The average temperature reactivity coefficient value is approximately -4.4395×10-5Δk/k per °C. The effect of thermal expansion has been included in the calculation of this temperature reactivity coefficient value. The average value of the void reactivity coefficient is +1.2606×10-3Δk/k per % of void fraction for the void fraction value range up to 50 %. However, molten salt fuel suffers only small changes in density during operation. The negative value of the temperature reactivity coefficient will be dominant; thus, the GAMA-MSR becomes inherently safe. The GAMA-MSR is designed for a 30-year operation time. The effect of the fission product to the reactor reactivity is -1.4597×10-2Δk/k. Full draining of the fuel to primary heat exchanger brings the reactor to shutdown condition with the reactivity of −0.20525. Results showed that this reactor has an inherently safe characteristic and has an effective shutdown system, especially due to the fuel draining system.

Utilizing even Plutonium Isotopes as burnable absorbers for controlling the reactivity and power distribution in Pressurized Water Reactors

In this paper, a new approach is proposed for controlling the reactivity and power distribution in pressurized water reactor cores without the implementation of burnable absorbers. This approach depended on the homogeneous mixing of uranium oxide with even plutonium isotopes; Pu-238, Pu-240, and Pu-242. Three Plutonium fuels; (99% UO2 + 1% Pu-238), (99% UO2 + 1% Pu-240), and (98% UO2 + 0.5% Pu-238 + 1% Pu-240 + 0.5% Pu-242) were simulated using the MCNPX 2.7 code. The three plutonium fuels are compared with the other three erbium fuels; (99.9% UO2 + 0.1% 167Er2O3), (99.74% UO2 + 0.26% 167Er2O3), and (99.61% UO2 + 0.39% 167Er2O3), which are taken as reference fuels. The weight percent of UO2, plutonium isotopes and enriched erbium is adjusted to provide the same initial reactivity. This is important for holding a fair comparison between the two fuel types. The simulation demonstrated that even plutonium isotopes, like enriched erbium, can improve reactor safety parameters. It was found that Pu-240 added fuel can suppress the initial reactivity by 27% and extend the fuel cycle by 6.9% compared to the conventional UO2. The reason for fuel cycle extension is due to the production of Pu-241. Meanwhile, Pu-238 and Pu-even added fuels provide suppressing the initial keff by 13% and 35% and shortening the fuel cycle by 2.7% and 21%, respectively compared to the reference UO2 fuel. Moreover, the fuel temperature coefficient (FTC), moderator temperature coefficient (MTC) and void reactivity coefficient (VRC) of Plutonium fuels are found to more negative than those of erbium fuels at the later stages of burnup because of the spectrum hardening resulting from Pu-239 production.

Characterization of neutronic parameters and radiation shielding design for an aqueous homogeneous reactor

In this research, an aqueous homogeneous reactor is simulated and its neutronic parameters are investigated. Due to the negative reactivity caused by temperature change and void formation, the cold excess reactivity is considered equal to 4 βeff. Also, calculations of radiation shielding and neutronic characterization of the reactor are performed. For different heights of uranyl sulfate fuel, the effective multiplication factor (keff) of the reactor is calculated under cold zero power and hot full power conditions to assess the required excess reactivity. The amount of fuel burnup and changes in the keff are investigated for 500 days, finally 200 operating days are chosen as the working duration of the reactor. The control rods worth (CRW) for different diameters of boron carbid and reactor safety parameters including shot down margin (SDM) and safety reactivity factor (SRF), temperature and void coefficients of reactivity are analyzed. In order to design a suitable radiation shield for this reactor, two different shields are considered. The first shield is a combination of polyethylene thicknesses and barite concrete, and the second shield is made only of barite concrete. The total thickness of the first and second shield was 230 and 180 cm, respectively. Due to the problems of using polyethylene (e.g. problems related to machining, shield dimensions and low melting temperature), the second shield is chosen as the suitable radiation shield for this reactor. Using this radiation shield, the total dose rates of neutron and gamma rays satisfy the ICRP dose limits.

The effect of central flux trap on-line sampling on Tehran Research Reactor(TRR) core safety parameters

Introducing the Neutron Flux Trap Channel to research reactors is an effective method to increase the neutron flux utilization for conducting the irradiation experiment and radioisotope production. The maximum thermal neutron flux could be achieved due to increased moderation in the flux trap area surrounded by the fuel elements, while all safety parameters and criteria must be met during the operation of the reactor. Regarding the TRR operator’s report about induced positive reactivity due to the entrance of some samples during reactor operation, it is predicted that this region is over-moderated, and the moderator reactivity coefficient is positive locally. Hence, this study aims to analyze experimental observations and the reactor safety parameters during online sampling into the flux trap channel. In this analysis, TRR Core’s different operating cycles were studied, and static, kinetic, and dynamic parameters of the reactor core with or without a flux trap channel were evaluated. For this purpose, some experiments were conducted to determine the safety parameters resulting from the sample placement, and then calculation results were validated against the experimental ones. Results indicate that void and temperature reactivity coefficients in the flux trap channel are locally positive; for example, the temperature and void reactivity coefficients for operating cycles with 33 fuel assemblies are +3.60 pcm/°C and +18.39 pcm/%, respectively. If 56.7% void is created in this channel, about 1.080$ positive reactivity will be induced to the core.Finally, accident analysis was conducted during various scenarios of reactivity insertion of different samples, and the variation of reactor power, reactivity, and fuel and clad temperatures were estimated in different transients. It is indicated that the maximum fuel and cladding temperatures in the worst case will be 183 °C and 144 °C, respectively, less than the melting point of aluminum meat of fuel.

The effect B impurity and Xe intrusion in the graphite moderator on void reactivity feedback of thermal MSR

Thermal Molten Salt Reactor (MSR) massively utilizes graphite as both moderator and reflector. The molten fuel circulates around the primary loop and directly contacts with the graphite. Gaseous fission products, such as neutron poison Xe-135, are normally separated out from the core. However, deposition of Xe-135 in graphite could take place through direct Xe-135 intrusion into graphite pores, or salt intrusion that eventually undergoes fission in the graphite and leaves the Xe-135 inside. Another possible way of strong neutron absorber deposition in the graphite is boron impurity. Beside reducing the core excess reactivity, the existence of strong neutron absorbers in graphite could alter the core reactivity feedbacks. Thus, this study is meant to analyse the effects of Xe-135 and natural boron deposition in the graphite to the core reactivity feedbacks. The evaluations were done by monte Carlo calculations on their effects on void coefficient, fuel, and moderator temperature reactivity feedbacks. The introduction of void into under-moderated thermal MSR core generally produces positive void reactivity feedback. The deposition of strong neutron absorbers in the graphite decreases the void reactivity coefficient, yet still positive. Despite of core excess reactivity loss, the existence of strong neutron absorber in the graphite is favourable in safety point of view.

Comparative analysis of fuel burnup calculations of fourth-generation European fast reactors

Abstract The sustainability of nuclear energy implies the continuous research and development of fast reactor technology. This work represents a comparative analysis of fuel burnup calculations of three commercial-level fast spectrum concepts which are under research and development in Europe. The investigated designs are the European Sodium-cooled Fast Reactor, the European Lead-cooled Fast Reactor, and the European Gas-cooled Fast Reactor. MCNPX transport code was used to design three representative fuel assemblies of these fourth-generation concepts to analyze and compare their neutronic and safety parameters. The neutronic and safety analysis in this work includes the evolution of infinite multiplication factor and fissile inventory, neutron yield, the average energy of neutrons causing fission, neutron energy spectrum, neutron generation time, Doppler reactivity effect, effective delayed neutron fraction, and coolant void reactivity coefficient. The fuel burnup results showed that the ESFR is superior regarding the reactivity swing and breeding of fissile 239Pu. On the other hand, analysis of safety parameters results showed that they depend mainly on fuel composition rather than on other design specifications.

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.

Whole core analysis of the single-fluid double-zone heavy water-moderated molten salt reactor (SD-HWMSR)

Thermal-spectrum molten salt reactor (MSR) concepts usually adopt graphite as a neutron moderator. Although graphite as a moderator has many advantages, it also has drawbacks, including a relatively short lifespan (it has to be replaced), positive temperature feedback, and loss of impermeability due to expansion. Replacing graphite with heavy water in MSRs can effectively solve the problems introduced by graphite. The SD-HWMSR is a Single-fluid Double-zone Heavy Water-moderated Molten Salt Reactor. Optimization of SD-HWMSR’s fuel channel pitch and radius has been conducted in this work. As a result, the SD-HWMSR with a high breeding ratio (1.07465 ± 0.00060), low initial 233U loading (1.43 t), and negative temperature and void reactivity coefficients is put forward. The SERPENT-2 is used to analyze the neutronics parameters of the reactor design. The current work investigates the change in the multiplication factor, breeding ratio, and the accumulation of significant nuclides in the core. The suitable 232Th and 233U refill rates needed to maintain criticality and enable analysis of the whole core of SD-HMMSR are thoroughly determined in this study. Uranium-233 and thorium-232 both have a maximum refill rate of 3.49 and 3.20 kg/d, respectively. While the average refill rate for 233U and 232Th during the 60 years of operation is 1.90 and 2.35 kg/d, respectively. The net production of 233U rises with time, and by the end of the 60 years, it is around 1.98 t. The doubling time for the SD-HWMSR is 31 years, which is consistent with previous results.

Investigation on Neutronic Parameters of the KLT-40S Reactor Core with U3Si2-FeCrAl using SCALE Code

From a safety point of view, the fuel-cladding of the current design of the KLT-40S reactor still carries a potential risk in the event of a loss-of-coolant accident (LOCA) allowing the formation of hydrogen gas. The concept of accident tolerant fuels (ATF) offers a variety of new safer fuel-cladding materials, one of which is U3Si2-FeCrAl, a potential fuel-cladding combination according to various research sources. In this research, a study of neutronic parameters (1) cycle length, (2) reactivity feedback coefficient, and (3) reactor proliferation resistance was performed with ATF material U3Si2-FeCrAl as fuel-cladding in the KLT-40S reactor core. Modeling and simulation of the ATF-fueled KLT-40S reactor core were performed using KENO-VI and TRITON modules from SCALE code. The results showed that replacement of the fuel-cladding material with the ATF material in the KLT-40S reactor resulted in a shorter cycle length, and the enrichment required to reproduce the original cycle length was above the safeguard limit. The fuel temperature, moderator temperature, and void reactivity coefficient were negative, although not as negative as the original ones. The spent fuel produced at the end of the cycle had good proliferation resistance, although not as good as the original one.

Neutronics simulation of China Experimental Fast Reactor start-up tests using FARCOB and ERANOS 2.1 code systems

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Prevalence and factors associated with sarcopenia in a cohort of patients with chronic pancreatitis

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Neutronic design of a PWR fuel assembly with accident tolerant-composite for the long-life core

For the future of nuclear power, the design and development of an economical, accident tolerant fuel (ATF) for use in the current pressurized water reactors (PWRs) are highly desirable and essential. It is reported that the composite fuels are advantageous over the conventional UO2 fuel due to their higher thermal conductivities and higher uranium densities. Due to higher uranium densities of the composite fuels, the use of composite fuels would lead to the significant increase of discharged burnup, thereby enhancing fuel cycle economy compared to that of the UO2 fuel. The higher thermal conductivities of composite fuels will increase the fuel safety margins. For implementation of the accident tolerant fuel concept, this study also investigates on the replacement of the conventional Zircaloy-4 cladding with SiC to minimize the hydrogen production due to interaction of water with cladding at high temperatures. In the present work, neutronic investigation of the composite fuels for a PWR has been conducted in comparison with that of the conventional UO2 fuel. Numerical calculations have been performed based on a lattice model using the SRAC2006 system code and JENDL-4.0 data library. Various parameters have been surveyed for designing a fuel with the UO2 and composite fuels such as U-235 enrichment, fuel pin pitch. In order to reduce the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. The temperature coefficients including fuel, coolant temperature reactivity coefficients, and both the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/t burnups without compromising the safety parameters including that four reactivity coefficients are considered those associated with the fuel temperature, coolant temperature, small (5%) void and large (90%) void. Further analysis of the performance of the UO2 and composite fuels in a full core model of a PWR is being conducted.

Void effect research on MOX-fueled lead-cooled fast reactor

Liquid metal-cooled reactor (LMR), including lead-cooled fast reactor (LFR), may have strong positive void reactivity coefficient due to the hard neutron spectrum, which is unfavorable for reactor reactivity control. This paper presents a detailed research on the lead void effect for deeper understanding on the causes. Through analytical analysis and numerical simulation, the lead void effect was decomposed into several components, and the spatial distribution of lead void worth was also calculated. The calculation results have shown that spectral hardening effect, which will cause an increase in fertile fission (mainly 238U) and decrease in the capture rate of all heavy nuclides, accounts for the high positive lead void reactivity. Meanwhile, it has been further concluded that high positive lead void reactivity mainly exists in active region. Finally, comparison between two different lead coolant lost scenarios were carried out to further prove these conclusions.

Model Comparison of Passive Compact-Molten Salt Reactor Neutronic Design Using MCNP6 and Serpent-2

Passive Compact Molten Salt Reactor (PCMSR) is a thermal breeder molten salt reactor (MSR) developed in Universitas Gadjah Mada, Indonesia, run in thorium fuel cycle. Its design was initially developed using deterministic code SRAC2006 but has never been compared with other codes. This paper attempts to compare PCMSR neutronic design using Monte Carlo codes MCNP6 and Serpent-2 with ENDF B/VII.0 continuous neutron cross-section library. The reactor was run in a pure thorium fuel cycle with lithium fluoride as its carrier salt. The analyzed parameters were effective multiplication factor (k eff ), temperature coefficient of reactivity (TCR), void coefficient of reactivity (VCR), and conversion ratio (CR). The result shows that there are several important discrepancies between the original calculation and this research. The Monte Carlo calculations implied that PCMSR core was able to be critical using lower fissile concentration than previously designed, but failed to reach CR above unity. While the TCR value was found to be negative, the VCR value was positive up until the 10 % void fraction. The PCMSR core suffered from ineffective neutron moderation and high neutron leakage. These findings imply that the previous PCMSR neutronic design is inaccurate. For PCMSR to be able to operate as a thermal breeder MSR, geometrical modifications must be performed to improve neutron moderation and reduce neutron leakage.

Neutronic Analysis of LEU-started Molten Chloride Fast Reactor without Fuel Reprocessing

One of the rarely explored molten salt reactor (MSR) designs is the molten chloride fast reactor (MCFR). This MSR design employs chloride salt instead of fluoride and operated in a fast spectrum. MCFR brings all the advantages of an MSR including breeding whilst being able to burn plutonium and minor actinides efficiently. Since not many countries have access to civilian plutonium, MCFR can also be started using low-enriched uranium (LEU). This study is an initial neutronic analysis of an MCFR using LEU as its startup fuel. Parameters analyzed are conversion ratio (CR) and its neutronic safety, namely effective delayed neutron fraction (βeff), temperature coefficient of reactivity (TCR), and void coefficient of reactivity (VCR). The core is divided into Core Zone and Blanket Zone. The fuel composition of NaCl-UCl3 with a molar fraction ratio of 60:40 and 50:50 is used in Core Zone and Blanket Zone, respectively. The neutronic calculation is performed using MCNP6 code with ENDF/B-VII library. For reference geometry, CR is valued at 0.9298, βeff at 0.00731, TCR at -19.8 pcm/°C, and average VCR at -154.31 pcm/void%. Thereby, the MCFR fulfills inherent safety criteria. Although its value is remarkably high, CR can be further optimized by modifying the separator and reflector material.

Neutronic analysis of fuel pin design for the long-life core in a pressurized water reactor

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.

Startup Test Plan and Predictions for Highly Enriched Uranium to Low-Enriched Uranium Fuel Conversion at the University of Missouri Research Reactor

Nonpower reactors licensed by the U.S. Nuclear Regulatory Commission require a startup test plan as part of any facility modification to verify operability prior to resumption of operations. In order to support conversion of the University of Missouri Research Reactor from the use of highly enriched uranium to low-enriched uranium (LEU) fuel, a startup test plan has been devised to measure certain reactor physics parameters for the initial all-fresh LEU core licensing documentation that will be submitted. These parameters include the approach to critical, primary coolant void coefficient of reactivity, flux trap void coefficient of reactivity, determination of flux trap sample reactivity worth, radial and axial thermal neutron flux mapping, control blade worth calibration, primary and pool coolant temperature coefficient of reactivity, and flux mapping of experimental positions. In this paper, predictions for these parameters made using the Monte Carlo N-Particle Version 5 (MCNP5) radiation transport code are reported. These predictions will support the startup tests by providing a baseline set of expectations and additional insight into the performance of the LEU core.

CROSSER - ПРОГРАММНЫЙ МОДУЛЬ ПОДГОТОВКИ ГРУППОВЫХ КОНСТАНТ ДЛЯ ИНЖЕНЕРНЫХ РАСЧЕТОВ БЫСТРЫХ РЕАКТОРОВ

The purpose of this work was to test the CROSSER high-speed constant preparation module for use in the practice of engineering calculations of the neutron-physical characteristics of fast reactors. Computational models of fast reactors BN-600, BN-800 and BN-1200 were used as test tasks. The following neutron-physical parameters were calculated: the criticality of the reactor model, the efficiency of the CPS and the sodium void coefficient of reactivity. The main reactor functionals were calculated using the MMKK engineering program, the constants for which were prepared using the CROSSER module. The initial sets of nuclear data for the MMKK program were the ABBN-93 library of constants and the constants in the ABBN-RFE formats obtained on the basis of the ROSFOND-2010 library. The results obtained using the MMKC program with detailed tracking of the particle energy were used as the reference calculation results. Detailed data in ACE format obtained from the ROSFOND 2010 library were used as the initial nuclear data for this program.

Neutronic evaluation of CANDU-6 core using reprocessed fuels

The spent fuel from a PWR still contains some amount of fissile materials depending on their initial enrichment and the burnup. Thus, spent fuel from PWRs containing about 1.5% of fissile material could be used as fuel for CANDU reactors after some fission products are removed from it. Thus, an important proposal is the DUPIC cycle, where spent fuels from a PWR are packaged into a CANDU fuel bundle using mechanical reprocessing but without the need of chemical reprocessing. When it is refueled with reprocessed fuel, the reactivity of the system increases, and this behavior may affect the safety parameters of the reactor. Therefore, this work studies the neutronic parameters of two reprocessing fuel techniques: AIROX and OREOX, which are evaluated for two different cores configuration. The first one considers heavy water as a moderator and coolant. The second one considers heavy water and light water as moderator and coolant respectively. These studies evaluate the core behavior based on the different number of reprocessed fuels channels and compare them with the reference core. To perform the simulation the MCNPX was used to calculate the effective multiplication factor, fuel temperature coefficient of reactivity, void reactivity coefficient, and neutron flux, which were evaluated at steady state condition for the different cases. The results show that the presence of parasitic absorbers in the reprocessed fuels hardens the neutron spectrum. This behavior provokes an increase in the core reactivity, in the fuel temperature coefficient and in the void reactivity coefficient. Among these parameters, the use of light water reduces the core reactivity but do not improve the fuel temperature coefficient and the void reactivity coefficient.