Power Coefficient Of Reactivity Research Articles

Physics Evaluation of Alternative Uranium-Based Oxy-Carbide Annular Fuel Concepts for Potential Use in Compact High-Temperature Gas-Cooled Reactors

Abstract Lattice physics calculations have been carried out to evaluate the performance and safety characteristics of a modified high temperature gas-cooled reactor (HTGR) prismatic fuel block concept, based on the MHTGR-350 benchmark problem. Key changes were to replace the conventional tri-structural isotropic (TRISO)-filled fuel compacts with heterogeneous, multilayer annular fuel pellets made with UCO, ThCO, or (U,Th)CO. These fuel pellets have multiple protective cladding layers of pyrolytic carbon and silicon carbide, which will give it robust qualities. With the increased loading of U-235 in the fuel block, it was necessary to replace up to 78 fuel holes and 42 coolant holes with a hydrogen-based moderator (7LiH), in order to ensure a thermal neutron energy spectrum in the lattice. Calculation results demonstrate that the modified fuel concept has several advantages and some challenges relative to the conventional MHTGR-350 design concept. With the increased uranium loading and the reduced neutron leakage due the use of 7LiH moderator rods, higher burnup levels and lower natural uranium consumption levels can be achieved with the same level of uranium enrichment. In addition, the expected fuel residence time increased by a factor of 20 or more, making such a concept very attractive for use in small, modular, “nuclear battery” HTGRs that would only need to be fueled once. Calculation results for the current concept indicate positive graphite and hydrogen moderator temperature coefficients, and further modifications will be required to ensure a negative power coefficient of reactivity.

A SEMI-ANALYTIC EIGENVALUE EXTENSION TO THE DOPPLER SLAB ANALYTIC BENCHMARK1

Advancement in multiphysics simulation has motivated interest in availability of analytic and semi-analytic benchmark solutions. These solutions are sought because they can be used to assess the accuracy of complicated numerical schemes necessary to simulate coupled physics systems. While there exist analytic solutions for fixed-source problems, benchmark-quality eigenvalue solutions are of interest because eigenvalue problems more closely align with analyses undertaken with coupled solvers. This paper extends a fixed-source benchmark, the Doppler Slab benchmark, to the eigenvalue case. A novel solution for this benchmark is derived. Numerical implementation of the benchmark is demonstrated through verification of numerical computation of the power reactivity coefficient.

A Co-ordinated Control Methodology for Rapid Load-Following Operation of a Pressurized Water Reactor Based Small Modular Reactor

The Pressurized Water Reactor (PWR) based Small Modular Reactors (SMRs) are best suited to perform load-following operation because of their inherent self-stabilizing and self-regulating characteristics. Therefore, these PWR based SMRs can not only acts as standalone electricity generating units but also work alongside intermittent renewable energy sources in Hybrid Energy Systems (HES) configuration. In both the cases, the PWR based SMRs are expected to perform rapid load-following operation along with satisfaction of constraints imposed by overall power coefficients of reactivity. The conventional mode of PWR operation-namely, the Normal Mode and Alternate Mode could not guarantee very rapid power maneuvering alongside satisfaction of the reactivity constraints. This paper puts forward a novel control methodology, which combines the advantages of both the Normal Mode and Alternate Mode, to accomplish very rapid power maneuver without defying the reactivity constraints. The load-following controllers were designed using Incremental Nonlinear Dynamic Inversion (INDI) which guarantees efficient tracking and does not require additional state observer and knowledge of inherent system nonlinearities. However, the nonlinear tracking controllers whose reference signal changes very rapidly are victims of unwanted glitches/spikes in control output which may ultimately lead to actuator wear-out. This problem has been handled by the use of a Reference Shaper whose detailed design methodology has also been presented. The performance of the control methodology and the controllers are verified in real-time using real-time Hardware-in-Loop (HiL) simulations.

Full compensation of reactive power in electric networks 0.4-10kV

The most important task in electric networks is to reduce energy losses. According to Network Company, in distribution electric networks with a voltage of 0.4-10 kV, they account for 62% of the total number of losses. One way to save energy is reactive power compensation. From Order of the Ministry of Energy of the Russian Federation dated June 23, 2015 No. 380 “On the Procedure for Calculation of the Ratio of Consumption of Active and Reactive Power for Individual Energy Receiving Devices (Groups of Energy Receiving Devices) of Electric Power Consumers” at a voltage of 0.4 kV, the reactive power coefficient should not exceed the value 0.35, and on the side of 10 kV - values of 0.4. In practice, in distribution electric networks, reactive power compensation devices are not installed in most cases, and the reactive power factor is 2-3 times higher than the standard values. Since compensating devices are expensive equipment, it is necessary to evaluate the feasibility of deep reactive power compensation, in the range of reactive power factor changes from its standard value to zero. The article discusses the economic and technical feasibility of deep reactive power compensation. The use of a generalized parameter of the electric network for determining the voltage at substations of distribution electric networks is proposed. The need for the coordinated use of reactive power compensation devices and booster transformers to ensure the quality of electricity is shown.

Conceptual design of a small-pressurized water reactor using the AP1000 fuel assembly design

This paper presents the conceptual design of a 300 MWt small modular reactor (SMR)using fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system and the JENDL-4.0 data library. The analysis showed that Doppler, moderator temperature, void, and power reactivity coefficients are all negative over the core lifetime. Semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel andclad surface temperature. The minimum departure from nucleate boiling ratio (MDNBR) is also calculated. The results indicate that a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

Анализ качества электроэнергии при стабилизации активной мощности дуговой сталеплавильной печи с помощью статического тиристорного компенсатора

The paper presents the key findings of studying the quality of electricity at the buses of an electric steel-making complex as affected by a static var compensator (SVC) running a novel control algorithm designed to improve the electric performance of an electric arc furnace (EAF) by stabilizing its active power at an optimal value. Analysis of how an SVC runs this algorithm reveals a considerable degradation of voltage quality compared to experiments with conventional algorithms. For detailed evaluation of SVC effects on the quality of electricity (QE), the research team ran an SVC in active-power mode and computed the QE values characterizing slow voltage deviations, non-sinusoidality, imbalance, and fluctuation; the values were compared against those attained by the reactive power- or voltage-controlled compensator. Experimentation used the mathematical model of an existing EAF-120-SVC unit, which was developed and run in MATLAB Simulink. To enable the SVC to sustain a constant active EAF power while also keeping the QE values sufficiently high, the paper proposes a structure of additional voltage and reactive power coefficient constraints, voltage balancing, and flicker suppression.

Measurement of Feedback Reactivity Effects of the Baec Triga Reactor

The aim of this study is to measure the feedback reactivity effects of the Bangladesh Atomic Energy Commission (BAEC) TRIGA Research Reactor (BTRR). Using digital instrumentation and control (I&C) system, feedback parameters such as reactivity coefficients of fuel, moderator and power effects are measured. At 300 kW reactor power, the obtained results for fuel temperature coefficient are 0.664 ꬿ/0c and 0.765 ꬿ/0c at C1 and D3 core positions respectively. The results for 100 kW to 400 kW reactor power show its increasing nature as the reactor power increases. The measured value of the moderator temperature coefficient is found to be 0.419 ꬿ/0c at 500 kW reactor power. The power coefficient of reactivity is measured for several power ranges from 50 kW to 2.5 MW and the average value is found to be 0.0757 ꬿ/kw. The measured values of the reactivity coefficients are found to be in a good agreement with the GA Safety Analysis Report (SAR) provided for BTRR.
 Journal of Bangladesh Academy of Sciences, Vol. 42, No. 2, 183-190, 2018

Power coefficient of reactivity: definition, interconnection with other coefficients of reactivity, evaluation of results of transients in power nuclear reactors

It is assumed by the authors of the present paper that with growing contribution of nuclear power in the production of electricity, nuclear power plants will be used to a higher degree in a manoeuvrable mode of operation rather than in the base-load mode. In other words, change of power from the nominal level to that of coverage of auxiliary loads will be becoming quite common and not so rare event as scheduled reactor shutdowns for fuel reloading or preventive works. There exist well-known problems in the use of nuclear reactors in the manoeuvrable operation mode, which include the task shared by all types of nuclear reactors. It is advisable to have a unified indicator weakly power-dependent and fairly easy to measure, which would make it possible to formulate the judgement about the nature of the transient processes within the entire power range and to assess the reactivity required for changing the power level by the preset value. Power reactivity coefficient (PRC) can be used as such indicator. Analysis was made of existing definitions and understanding of PRC in relevant references. It turned out that there is no generally accepted definition of the PRC. Based on the performed study, the following definition was suggested: the PRC is the ratio of the low reactivity introduced into the reactor to the power increment at the end of the transient process. It is assumed here that variation of reactivity is dependent on the energy released in nuclear fission but is not related to the changes of reactivity induced by feedback signals in the automatic reactor power control system. Analysis of the relationship between the PRC and temperature coefficients and technological parameters associated with the steady-state control program was performed taking the above suggested definition into account. PRC calculations were performed using the simplest model of VVER-1000 type power reactor. It was found that PRC is weakly power-dependent. The purpose of the present study is to investigate dependence of PRC on the temperature reactivity effects and on the technological parameters associated with the steady-state control program of the power unit, using the example of VVER-1000. Effects of PRC on the static and dynamic power reactor operation modes are analyzed.

Development of methodology for efficient fuel design evaluation of the Advanced High Temperature Reactor (AHTR)

This paper presents a methodology for developing a neutronics model to be used for a fluoride-salt high temperature reactor (FHR) fuel design optimization study. The reactor design is based on the Advanced High-Temperature Reactor (AHTR) with hexagonal fuel element with fuel “planks” (plates) developed at the Oak Ridge National Laboratory. The overall objective of this research was to develop a methodology accurate enough, yet practical and fast enough for a fuel plank design optimum search that minimizes the fuel cycle cost (FCC) of FHR. Resonance self-shielding calculations for double heterogeneous fuel planks of AHTR design is currently unavailable in most codes. Therefore, automated modeling of the double heterogeneity was implemented to allow accurate use of multi-group simulations. The methodology generates MCDancoff Factors using SCALE6.2; its adequate accuracy was verified. A MCDancoff Factor surrogate model was created so on-the-fly MCDancoff Factors could be provided for depletion calculations over the design space for parametric studies. A simple non-linear reactivity model with assembly-to-whole-core reactivity correction was applied to analyze performance of multi-batch refueling. Results indicate that cycle length is maximized at 40% packing fraction with a carbon-to-heavy metal ratio (CHM) of approximately 250, which minimizes the cost of outages. In contrast, discharge burnup and fuel utilization is maximized at around 20% packing fraction and a CHM of approximately 600, which minimizes fuel cost, but not necessarily the FCC cost, which also accounts for outage penalty. Void and power reactivity coefficients were determined over the design phase space to establish the envelope for stable design. The minimum FCC is the optimum trade-off between the two, within the limits of the stability envelope. In combination with the cost model the objective function is non-linear and requires global optimization methods. This paper presents the methodology needed for global optimization, while the optimization itself will be presented in a follow up companion paper.

Simulations of Pressure-Tube–Heavy-Water Reactor Cores Fueled with Thorium-Based Mixed-Oxide Fuels

Thorium, a fertile nuclear fuel which is nearly three times as abundant as uranium, represents a long-term energy resource that could complement uranium by making the current nuclear fuel cycle more sustainable. With the expected refurbishment and new construction of pressure-tube–heavy-water reactors (PT-HWRs) within the international community, there is a good opportunity to gain experience with thorium-based fuels and to start the transition phase toward use of thorium as part of the nuclear fuel cycle. Previous studies have shown that in the near term (10 to 15 years), small amounts of thorium could be introduced into conventional 37-element fuel bundles made with natural uranium (NU) or slightly enriched uranium for use in PT-HWRs, which could help improve performance and safety characteristics relative to the use of NU alone. This work is an evaluation of various types of thorium-based oxide fuels in 35-element bundles that could potentially be implemented in current PT-HWRs to enable the transition toward the thorium fuel cycle to harness the energy potential from thorium and to conserve uranium resources. Such fuel concepts could have additional benefits, including enhanced reactor performance and safety characteristics, a reduction in the production of fissile plutonium, and the ability to consume stockpiles of plutonium. Thorium (in the form of ThO2) is mixed with a fissile driver of either reactor-grade plutonium (67 wt% Pu-fissile/Pu, 3.5 to 4.5 wt% PuO2), low-enriched uranium (LEU) (5 wt% 235U/U, 40 to 50 wt% LEUO2) or 233U (1.8 wt% 233UO2). The 35-element fuel bundle has a central graphite displacer rod to help maintain a low coolant void reactivity (CVR), while also flattening the radial power distribution in the fuel bundle due to supplementary neutron moderation. Analysis suggests that these thorium-based fuel types have lower fuel temperature coefficients (FTCs) and CVR values and higher fissile utilization (FU) compared to conventional 37-element natural uranium oxide (NUO2) fuel used currently in PT-HWRs. Lower values of CVR and FTCs will lead to lower and more negative values of the power coefficient of reactivity (PCR), which will help increase the safety margin for PT-HWRs further. Achieving higher FU will help better conserve nuclear fuel resources (both uranium and thorium), and should lead to lower fuel costs in a once-through fuel cycle. The use of a higher–fissile content fuel (in the form Pu, LEU, or 233U mixed with thorium) enables both higher fuel burnups and FU relative to NU, while also breeding and burning 233U. The low-burnup (LEU,Th)O2 core configuration shows the most promise as a feasible and practical means of producing energy from thorium (up to 33% of total energy produced) while also generating stockpiles of 233U. PT-HWR cores using (Pu,Th)O2 fuels may have some challenges in regard to the power balance in the bundle and reactor operation. Without reactivity control devices, some (Pu,Th)O2 cores may require power reductions to avoid exceeding bundle power limits, if a maximum linear element rating of 54 kW/m (which applies to uranium-based fuels) is assumed. PT-HWR cores fueled with (Pu,Th)O2 also demonstrate significant benefits as a means of consuming stockpiles of fissile plutonium, reducing inventories by 84% to 90% in a single pass through a PT-HWR core, while also producing substantial quantities of 233U (up to 314 kg/year/reactor) in the spent fuel for potential reprocessing and recycling in a closed fuel cycle.

Мощностной коэффициент реактивности: определение, связь с коэффициентами реактивности, оценка результатов переходных процессов энергетических реакторов

Мощностной коэффициент реактивности: определение, связь с коэффициентами реактивности, оценка результатов переходных процессов энергетических реакторов

A novel approach for bifurcation analysis of out of phase xenon oscillations using multipoint reactor kinetics

In this paper, an approach for carrying out bifurcation analysis of out of phase xenon oscillations has been demonstrated using a reduced order model. The reduced order model consists of multipoint kinetic equations coupled with xenon and iodine equations along with a lumped power coefficient feedback. A large Pressurised Heavy Water Reactor (PHWR) has been considered. The reactor core is divided into two regions for the analysis to present a “proof of principle” for this approach. However, extension of this approach for multiple regions, for better accuracy, though not trivial is quite straight forward. Bifurcation analysis has been carried out using nonlinear analysis tools like MATCONT and BIFOR considering the operating power level of operation, power coefficient of reactivity and reactor size as bifurcation parameters. The results show the existence of Generalized Hopf (GH) points which divides the parameter space into subcritical and supercritical Hopf bifurcation regions. The identification of subcritical Hopf bifurcation is important as it shows unstable limit cycles in the region identified as stable by linear stability analysis.

Measurement and Analysis for Determination of PCR of the CANDU6 Core

The power coefficient of reactivity (PCR) needs to be negative to achieve the inherent safety of a reactor. However, the possibility that the PCR of CANada Deuterium Uranium (CANDU) reactors can be positive has been raised in recent studies. In such circumstances, there was an experimental approach on evaluating the PCR of CANDU in 2012 at an in-operation CANDU reactor, Wolsong Unit 2. In the evaluation, the PCR was indirectly measured by a method that required estimating the reactivity variation due to Xe, liquid zone controllers (LZCs), and fuel depletion based on the measurement data. In this study, the PCR of a CANDU was reevaluated by the same methodology with more proper and detailed methods to estimate all the factors in addition to some minor reactivity corrections. The estimation of Xe and LZC reactivity was performed by an in-house three-dimensional code and Serpent2 in addition to RFSP-IST. Furthermore, several short studies regarding the factors that result in uncertainty of the Xe/LZC reactivity estimation were done in detail. First, a method to determine 14 LZC levels at a certain time based on the measurement data was appropriately selected through determining the features of the measurement data. The influence of the power transient scheme and the impact of local refueling transients due to daily refueling of CANDU reactors on xenon reactivity estimation were also analyzed briefly. Finally, the PCR of the CANDU in operational conditions was evaluated to be ~0.5 pcm/%P on average at a measurement time of 5 to 20 min after the power perturbation.

Use of Er Burnable Absorber for Improvement of CANDU6 Safety Parameters

This paper reports on the improvement of the power coefficient of reactivity (PCR) and minimization of the coolant void reactivity (CVR) of a CANDU6 reactor. A burnable absorber of Er2O3 (erbia) was mixed homogeneously with UO2 fuel in the central fuel element to maximize the Doppler broadening and minimize the CVR of the CANDU6 reactor. In this study, recovered uranium (RU) with 0.9 wt% 235U enrichment was utilized in the advanced CANFLEX fuel bundle instead of natural uranium (NU). First, the optimal loading of erbia was investigated through lattice-based analysis, and its impact on the lattice characteristics was examined. In particular, both the fuel Doppler effect and CVR were evaluated for the RU-loaded lattice. For a more reliable analysis, a three-dimensional (3-D) equilibrium core was determined based on the standard time-average methods for erbia-loaded CANDU6 cores using the Serpent-COREDAX/CANDU code system. The core analysis was based on a hybrid two-step method in which the lattice analysis was performed by the Serpent Monte Carlo code, and the 3-D whole-core analysis was done using a diffusion theory–based nodal code named COREDAX. For the derived equilibrium cores, the core performances were evaluated in terms of the fuel burnup and power profile. Additionally, the safety parameters, including the PCR and CVR, were evaluated for the equilibrium core conditions. The safety parameters of the 3-D whole core were compared with those obtained with simple lattice-based analysis. It was observed in the analysis that Er-loaded CANFLEX-RU fuel provides a 60% more negative fuel temperature coefficient than standard CANDU-NU fuel.

An Improved Pilot Relaying Scheme for Shunt Compensated Transmission Line Protection Based on Superimposed Reactive Power Coefficients

The fault component power and energy based relaying schemes have limitation during double-phase to ground fault due to the mutual coupling effect. This article presents a superimposed reactive power coefficient (SRPC) based pilot relaying scheme for the protection of shunt compensated transmission line. SRPC of a phase is defined as the ratio of integrated superimposed reactive power (SRP) of that phase to the maximum magnitude of integrated SRPs among all three phases. If SRPC of any phase is found less than −0.5, an internal fault is detected and that phase is considered as a faulty phase. If no phase possesses SRPC less than −0.5, the fault is an external fault. The algorithm will start only if the superimposed differential current of any phase is found more than 50% of pre-fault differential current. The results are obtained using PSCAD/EMTDC software for different conditions, such as different fault resistances, fault locations, static VAR compensator locations, and source impedances. The proposed scheme is found robust and accurate for all the cases considered. The performance of the proposed scheme is superior to the recently proposed similar type of scheme and is unaffected from capacitive coupled voltage transformer and current transformer measurement errors and synchronization error.

An Interval Approach to Nonlinear Controller Design for Load-Following Operation of a Small Modular Pressurized Water Reactor

A pressurized water reactor (PWR) is most suitable for load-following operation because of its self-regulating nature. However, rapid power maneuvering at an enhanced rate is heavily constrained by the overall power coefficient of reactivity which depends on both the reactor neutronics and thermal hydraulics. In this paper, a 100 MWe PWR module with an integrated steam generator, is modeled using the nonlinear equations representing both the neutronics and thermal hydraulics. Such a PWR is typically an integrated structure with a steam generator included in the same casing and can either be connected to the grid or connected with other such modules using a multimodular approach to constitute a Nuclear Steam Supply System (NSSS) driving a common turbine and an alternator. In this paper, Nonlinear Dynamic Inversion (NDI) technique is used to design a controller capable of controlling such a reactor for load-following with frequent changes in demand, in both the modes of operation, namely, reactor-follow-turbine and turbine-follow-reactor. With the assumption that the parameters of the reactor vary within intervals as the power varies, an interval approach coupled with NDI ensures that the controller satisfies the operating constraints. The methodology is established for both the modes with credible real-time Hardware-in-Loop (HiL) simulations.

Energy upgrading of buildings. A holistic approach for the Natural History Museum of Crete, Greece

This article presents the fundamental points of the accomplished work regarding the energy upgrading of the building of the Natural History Museum of Crete, Greece, based on a cluster of passive and active measures. The main target was the building's energy efficiency upgrading from D to A+ rank. Among the passive measures, insulation of the building's envelope, installation of new windows and doors, construction of a green roof and planting of an outdoor space at the south side of the building to improve the ambient climate locally are included. Furthermore, small wind turbines and a photovoltaic station on the roof, geothermal heat pumps with an open loop system operating with seawater, new lighting equipment controlled by a central management system and reactive power coefficient compensation constitute the proposed active systems. Energy saving percentages from 40% to 93% are achieved with all the proposed technologies. In total, the primary energy specific consumption is reduced from 273.65 kWh/m2 at 18.36 kWh/m2, giving a total energy saving percentage of 93.29%. The building's energy efficiency is upgraded from the rank D to the rank A+, according to the European Union's standards. The total cost of the proposed measures is estimated approximately at 900,000€.

Measurement of the Power Coefficient of Reactivity of VVR-Ts in a Wide Power Range

The power coefficient of reactivity of VVR-Ts was previously determined only in the range 10–11 MW. The aim of the present work was to determine the power coefficient of reactivity in the range 1–10 MW. The experiment was performed on a completely depoisoned reactor. The change in the reactivity excess was determined according to the position of the control rods. It was shown experimentally that the power coefficient of reactivity is not constant and is negative for the given power range. Thus, the VVR-Ts reactor can operate safely in different operating regimes.

Neutronics and thermal hydraulics analysis of a small modular reactor

The small modular reactor (SMR) offers many feasible pathways for the construction of more nuclear power plants. A physics model of a near term deployable SMR of the integral pressurized water reactor (IPWR) design is developed. Fuel depletion simulations are performed to optimize the active fuel length, fuel enrichment and core loading pattern in order to achieve a uniform core power distribution. The optimized core can produce 500MW of thermal power with a four year core life-time at a capacity factor of 87%. The core consists of 69 uranium dioxide (UO2) fuel assemblies; 5 assemblies at 4.4at% 235U enrichment and 64 assemblies at 4.95at% 235U enrichment. The active fuel length is 200cm and the core diameter is 194.55cm for an active core height-to-diameter ratio of 1.03. As part of the study the active fuel length is increased to 240cm resulting in an increased capacity factor of 95% at 530MW of thermal power output for an active core height-to-diameter ratio of 1.23. Rod cluster control assemblies (RCCAs) are placed strategically to reduce the overall core power peaking factor to 1.3. Estimated reactor kinetics parameters such as the delayed neutron fraction and mean neutron generation time are typical of existing larger pressurized water reactors (PWRs) from which much of the IPWR based SMR design is derived. This study showed that Doppler, moderator temperature, void and power reactivity coefficients are all negative over the core life-time of four years indicating the possibility of safe reactor operation. A semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin from industry standards on peak fuel and clad surface temperature limits. The critical heat flux (CHF) is calculated and is not exceeded even in 10% overpower conditions. In addition the nucleate boiling ratio (DNBR) is calculated and found to be above 4.8 for the entirety of the active core region. These parameters further engender confidence in the safety of the SMR design.

A Lattice-Based Monte Carlo Evaluation of Canada Deuterium Uranium-6 Safety Parameters

Abstract Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k -infinity values is only 0.5–1 pcm.