Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O_2−x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of “best estimates” provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.     

ECRIX-H experiment: Synthesis of post-irradiation examinations and simulations

The purpose of the ECRIX-H experiment is to study the behaviour of a composite ceramic target made of AmO_1.62 microdispersed in an MgO matrix irradiated for 318 EFPD in the Phenix sodium-cooled fast reactor (SFR), in a specific carrier sub-assembly equipped with annular blocks of CaH_x acting as a neutron moderator. Results indicate that magnesia-based inert matrix targets display satisfactory behaviour and moderate swelling under irradiation, even for significant quantities of helium produced and a high burn-up. On this basis, the design of transmutation fuel pins for recycling of minor actinides (MA) in accelerator-driven systems (ADS) or in fast neutron reactors (FR) could be optimised so as to increase their performance level (initial MA content, burn-up, etc.).The measured Am fission rate (25 at.%) was found to be lower than that predicted by neutronic simulations probably due to the inaccuracies linked to the complexity of neutron modelling and the uncertainties on nuclear data related to moderated neutron spectrum. In addition, as most of the initial Am transmuted into Pu under irradiation, a PuO_x-type phase was created within the initial AmO_1.62 particles, leading to the incomplete dissolution of the irradiated targets under standard reprocessing conditions. This issue will have to be considered and investigated in greater detail for all transmutation fuels and targets devoted to the multi-recycling of MA.     

Physical studies of minor actinide transmutation in the accelerator-driven subcritical system

The accelerator-driven subcritical system(ADS)with a hard neutron energy spectrum was used to study transmutation of minor actinides(MAs). The aim of the study was to improve the efficiency of MA transmutation while ensuring that variations in the effective multiplication factor(k_(eff)) remained within safe margins during reactor operation. All calculations were completed using code COUPLE3.0. The subcritical reactor was operated at a thermal power level of 800 MW, and a mixture of mononitrides of MAs and plutonium(Pu) was used as fuel.Zirconium nitride(ZrN) was used as an inert matrix in the fuel elements. The initial mass composition in terms of weight percentages in the heavy metal component(IHM)was 30.6% Pu/IHM and 69.4% MA/IHM. To verify the feasibility of this MA loading scheme, variations in k_(eff), the amplification factor of the core, maximum power density and the content of MAs and Pu were calculated over six refueling cycles. Each cycle was of 600 days duration, and therefore, there were 3600 effective full power days.Results demonstrated that the effective transmutation support ratio of MAs was approximately 28, and the ADS was able to efficiently transmute MAs. The changes in other physical parameters were also within their normal ranges.It is concluded that the proposed MA transmutation scheme for an ADS core is reasonable.     

Role of ADS in the back-end of the fuel cycle strategies and associated design activities: The case of Japan

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the “double-strata” fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead-bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.     

Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

In the frame of Partitioning and Transmutation (P&T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a parametric sensitivity study of both the fuel cycle and the electricity generating costs.     

Monte Carlo calculations on transmutation of plutonium and minor actinides of pebble bed high temperature reactor

The paper aimed to maximize the fuel burnup performance of plutonium and minor actinides fueled pebble bed high temperature reactor (PBMR-400). The PBMR-400 was designed as a reference core. The neutronic calculations were performed by the code combination MCNP-ORIGEN-MONTEBURNS. In this study, neutronic performances of three different types of nuclear fuels (Reactor Grade Plutonium – RGPu; Weapon Grade Plutonium – WGPu and Minor Actinides – MAs) combined with natural uranium were conducted in a PBMR-400 full core. The neutronic performances were compared with the original uranium fuel designed for this reactor. Neutronic calculations showed that 9.6 wt % enriched uranium has a core effective multiplication factor (k_eff) of 1.2395. Corresponding to this k_eff values the natural UO₂/RG-PuO₂; natural UO₂/WG-PuO₂ and natural UO₂/MAO₂ mixture were found 70%/30%, 76%/24% and 63%/37%, respectively. The operation times were computed as ∼2000, ∼2500 and 1400 days whereas, the corresponding burnup values were obtained as ∼163 000, ∼194 000 and ∼116 000 MWD/T, respectively, for end of life k_eff set equal to 1.08.     

Prompt and power reactivity coefficients for the next generation VVER-1000 reactor including hexagonal assemblies and annular fuels

Many neutronics as well as thermal-hydraulics calculations have been made to find the performance of the proposed annular fuels (internally and externally cooled fuel pins) for both next generation PWRs and BWRs. Specifically, there has not been a significant study on the Russian type VVER-1000 reactors with annular fuels. Our aim herein is to study two important safety coefficients of the Iranian VVER-1000 core including hexagonal annular fuel assemblies at its BOC. The safety coefficients are “prompt reactivity coefficient” and “power reactivity coefficient”, where all simulations are made using MCNP-5 code. We found less (absolutely) Doppler coefficient for the next generation VVER-1000 and therefore Doppler coefficient decreasing is a good feature to avoid more resonance neutrons absorbing in the U-238; causes more fission density and also less soluble boron for core controlling (at the BOC) with comparing to the current VVER-1000 solid pins.     

Core neutronics characterization of the GFR2400 Gas Cooled Fast Reactor

The Generation IV initiative was launched with the goal of developing nuclear reactors which surpass current designs in safety, sustainability, economics and non-proliferation. From the six most promising concepts the Gas Cooled Fast Reactor (GFR) represents a challenging and innovative idea that is prominent in the sustainability aspect with the ability to have a closed fuel cycle and the potential to burn minor actinides (MAs). The European FP7 GoFastR project was one of the latest steps in the development and further optimization of GFRs.This paper presents a comprehensive overview of the neutronic performance of GFR2400 which was considered as a conceptual design for a large scale GFR within the collaboration. This reactor is the newest on the evolutionary path of fully ceramic GFRs featuring ceramic fuel and structural materials allowing high temperatures and efficiency using helium coolant. An important innovation of the current design is the application of refractory metallic liners to enhance the fission product retention of the cladding, resulting in a significant neutronic penalty during normal operation, at the same time being advantageous under transient conditions involving spectrum softening.Using the ERANOS and SCALE code systems several parameters were determined for beginning of life (BOL) conditions, including excess reactivity, various reactivity effects such as depressurization, Doppler or thermal expansion effects, as well as kinetic parameters. An extensive sensitivity and uncertainty analysis of these parameters was also done with the 15 group BOLNA and 44 group SCALE covariance libraries. Open and closed fuel cycle operations were investigated and the transmutational capabilities were studied with the GFR connected to traditional light water reactors in a symbiotic system.The presented analysis shows that the GFR2400 design is a major improvement compared to previous concepts. All preliminary constraints are respected resulting in a manageable initial Pu inventory of 10 t/GWel at 45% plant efficiency, a low MA mass fraction of 1% by self-recycling and a near zero breeding gain without the use of fertile blankets. At the same time the reactor has acceptable safety features precluding super-prompt-criticality in depressurized conditions at BOL and in open cycle equilibrium. Either of the two planned control devices is sufficient to shut down the reactor independently of the other and the refractory liners introduce significant negative reactivity in case of water ingress. However the occurrence of hot spots when all control rods are inserted needs further analysis.The design also shows promising closed fuel cycle and transmutational performance. However – as is the case in other fast reactors – the fuel cycle closure causes safety related parameters to degrade, most importantly the depressurization reactivity effect to exceed the effective delayed neutron fraction in the current design. To assess the acceptability of this deterioration further analysis is needed.Finally, it can be concluded that current commercial codes are satisfactory for such analysis; however there is a need for better covariance data. Several parameters exceed their target uncertainty value, most notably the k-effective by a factor of 6, the main source of the uncertainty being the inelastic scattering of ²³⁸U.     

Partitioning and transmutation studies at JAERI both under OMEGA program and High-Intensity Proton Accelerator Project

At the Japan Atomic Energy Research Institute (JAERI), active and comprehensive studies on partitioning and transmutation (P&T) of long-lived nuclear waste from the reprocessing processes of spent fuel has been carried out under the OMEGA program. Studies at JAERI include a design study of dedicated transmutation systems both of an MA burner fast reactor (ABR) and an accelerator-driven subcritical system (ADS), and the development of a high intensity proton accelerator as well as the development of partitioning process, nitride fuel fabrication/dry separation process technologies and nuclear data studies.

During the course of studies, JAERI developed the concept of the double-strata fuel cycle, in which a dedicated system is used for transmutation. Comparing the various transmutation systems, such as thermal neutron spectrum or fast neutron spectrum systems, power reactors or dedicated systems, from the viewpoints of reactor physics, nuclear fuel cycle and socio-technical issues, it was concluded that the ADS is the best option for transmutation of minor actinide(MA). JAERI, therefore, decided to concentrate its R&D efforts on the development of ADS and related technologies.

One of the goals of R&D is to provide a basis for designing demonstration facilities of ADS, aqueous partitioning process and nitride fuel fabrication and dry separation technologies. As the initial step toward this purpose, the construction of an ADS experimental facility is planned under the High-Intensity Proton Accelerator Project which JAERI and the High Energy Accelerator Research Organization (KEK) are jointly proposing since 1998.

The paper discusses the some of the results of P&T studies and the outline of the High-Intensity Proton Accelerator Project under which ADS experimental facility will be constructed.     

Coolant void reactivity adjustments in advanced CANDU lattices using adjoint sensitivity technique

Coolant void reactivity (CVR) is an important factor in reactor accident analysis. Here we study the adjustments of CVR at beginning of burnup cycle (BOC) and k_eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice using the optimization and adjoint sensitivity techniques. The sensitivity coefficients are evaluated using the perturbation theory based on the integral neutron transport equations. The neutron and flux importance transport solutions are obtained by the method of cyclic characteristics (MOCC). Three sets of parameters for CVR-BOC and k_eff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR-BOC (CBCVR-BOC). To approximate the EOC sensitivity coefficient, we perform constant-power burnup/depletion calculations using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Our aim is to achieve a desired negative CVR-BOC of −2 mk and k_eff-EOC of 0.900 for the first two cases, and a CBCVR-BOC of −2 mk and k_eff-EOC of 0.900 for the last case. Sensitivity analyses of CVR and eigenvalue are also included in our study.     

The neutronic design of a critical lead reflected zero-power reference core for on-line subcriticality measurements in Accelerator Driven Systems

In the framework of the GUINEVERE project (Generation of Uninterrupted Intense NEutrons at the lead VEnus REactor) carried out at SCK⋅CEN in Mol, the continuous deuteron accelerator GENEPI-3C is coupled to the VENUS-F fast simulated lead-cooled reactor. With this installation the neutronic behaviour of an Accelerator Driven System (ADS) during different phases of operation is studied. Because of the subcritical character of ADS, an on-line reactivity monitoring system is required. The main goal of the GUINEVERE project is to design such a system, to evaluate its performance and accuracy, and to evaluate an extrapolation towards full-size ADS.     

Reactor physics ideas for large scale utilization of thorium in gas cooled reactors

The physics principles for maximizing the fertile to fissile conversion were used in developing reactor concepts for large scale utilization of thorium in thermal and fast reactors (Jagannathan & Pal, 2006; Jagannathan et al., 2008). It is recognized that these principles are very well suited for ‘He’ gas cooled reactors with graphite moderator since both helium gas coolant and the graphite moderator have low neutron absorption characteristics and thus gives better neutron economy. In this paper, these ideas are applied to the High Temperature Test Reactor (HTTR) core of Japan to assess its advantage over the present day gas cooled reactors. HTTR is helium cooled and graphite moderated system. Significant amount of thorium has been loaded in the HTTR core with some minimal changes in the existing core design. The modified design is called HTTR-M core.In the HTTR-M core, the fuel is changed from enriched UO₂ fuel to Pu in ThO₂ fuel. The locations of boron type burnable poison rods within each fuel assembly of HTTR are replaced by one cycle irradiated thoria rods. Also, the B₄C type control assembly around the HTTR core is replaced by fresh seedless thorium assembly. The fertile thoria assembly are scattered uniformly in the HTTR-M core. The equilibrium core of HTTR-M shows very small burnup reactivity swing. The core excess reactivity is ∼18 mk at BOC and reduces to 1 mk at 660 days. It is interesting to note that this small reactivity change is intrinsically achieved by the choice of seed and fertile dimensions and their contents without the use of burnable poison rods or mechanical control rods which are used in HTTR core. The burnup reactivity swing in the latter after using burnable poison is ∼100 mk. The fissile seed inventory ratio (FIR) in a fuel cycle is 0.90 as compared with 0.717 of HTTR core. Since ²³³U is a better fissile nuclide with highest ‘η’ value in thermal range, the above conversion ratio can be regarded as quite good.     

Determination of the subcriticality level using the Cf source-detector method

Measurement and monitoring of reactivity in a subcritical state, e.g. during the loading of a power reactor, has a clear safety relevance. The methods currently available for the measurement of k_eff in stationary subcritical conditions should be improved as they refer to the critical state. This is also very important in the framework of ADS (accelerator driven systems) where the measurement of a subcritical level without knowledge of the critical state is looked for. An alternative way to achieve this is by mean of the ²⁵²Cf source-detector method. The method makes use of three detectors inserted in the reactor: two “ordinary” neutron detectors and one ²⁵²Cf source-detector which contains a small amount of ²⁵²Cf that introduces neutrons in the system through spontaneous fission. By observing fissions through the detection system and correlating the signals of the three detectors, the reactivity ρ (and hence the multiplication factor k) can be determined.     

Minimization of an initial fast reactor uranium-plutonium load by using enriched lead-208 as a coolant

Long-term scenarios of nuclear energy evolution over the world scale predict deployment of fast reactors (FRs) from 2020 to 2030 and achievement on 2050 the world installed capacity equal to 1500 GW_e with essential increasing the FRs number. For several countries (i.e. Russia, Japan) whose policies are based on a sharp increase of nuclear production, at the stage near 2030-2040 when plutonium, Pu, from the PWR spent nuclear fuel is consumed, the Pu lack will stimulate minimization of its load in FRs. The period of Pu deficiency will be prolonged till the years when breeding gain (BG) equal to 0.2-0.3 in fast breeding reactors (FBRs) is obtained which corresponds to Pu inventory doubling time of 44-24 years.In this paper one of opportunities to minimize fuel loading is considered: it is related to using a low neutron capturing lead isotope, ²⁰⁸Pb, as a FR coolant. It is known, that natural lead, ^natPb, contains a stable lead isotope, ²⁰⁸Pb, having a small cross-section of neutron capture via (n, γ) reaction. In the paper it is shown that the macroscopic cross-sections 〈σ_n,γ〉 of radiation neutron capture by the lead isotope ²⁰⁸Pb averaged on the ADS core neutron spectra are by ∼3.7-4.5 times less than the corresponding macroscopic cross-sections for a natural mix of lead isotopes ^natPb. This circumstance allows minimizing load of a lead fast reactor (LFR) core for achievement its criticality, as well as the load of an accelerator-driven system (ADS) subcritical core—for achievement of its small subcriticality. In using ²⁰⁸Pb instead of ^natPb in the ADS blanket, the multiplication factor of the subsritical core, K_eff, could be increased from the initial value K_eff = 0.953 up to the value of K_eff = 0.970. To achieve this higher value of K_eff in the same core cooled by ^natPb an additional amount of 20-30% of U-Pu fuel will be needed.The isotope ²⁰⁸Pb content in the natural mix of isotopes, ^natPb, is high enough, above 52%, and its separation in large amounts (several tens’ and hundreds’ of tonnes) is expensive but really solvable technical task. In the project (ISTC #2573, 2005), developed with authors’ participation, it is shown that a new laser photochemical technique of lead isotope separation, being developed in future, permits to obtain large quantities of ²⁰⁸Pb under its acceptable price, of close to $200 kg⁻¹.     

Applicability of dynamic programming to the accelerator-driven system (ADS) fuel cycle shuffling scheme for minor actinide (MA) transmutation

The accelerator-driven system (ADS) is under development in several countries to reduce the burden for conditioning and disposal of the high-level radioactive waste (HLW) by transmuting minor actinide (MA). It is expected that the fuel shuffling can make the power distribution flat and transmute MA effectively by using only one kind of fuel composition. But the total number of calculation cases becomes huge for finding the globally optimum case. In order to find the best shuffling scheme for MA transmutation by ADS, we attempt to develop a calculation code within an acceptable time by employing dynamic programming. It is used successfully for a cylindrical core with three fuel regions with 20 times of fuel shuffling.     

Re-evaluation of the Effective Delayed Neutron Fraction Measured by the Substitution Technique for a Light Water Moderated Low-enriched Uranium Core

The effective delayed neutron fraction β_eff for a light water moderated low-enriched UO₂ core has been re-evaluated to obtain benchmark data for the validation of calculation codes and nuclear data. Originally, the β_eff value was measured by the substitution method. In that method, the β_eff value was obtained from measured reactivity change by substituting a Sb-Cd-Pb absorber rod for a 2.6 wt% UO₂ rod for all core regions. In the present evaluation, we have employed the latest value for the buckling coefficient of reactivity to re-evaluate the substitution reactivity with high accuracy. In addition, the correction factor, which was ignored in the previous measurement, has been calculated to compensate the difference in the absorption cross sections of fuel and absorber rods. Consequently, the obtained β_eff value in the present evaluation was 0.00771±0.00017, and it is more credible than the previous one. The present result is available as benchmark data for the verification of delayed neutron data for light water reactors.

For comparison, we have calculated the β_eff value using a transport code TWODANT with the JENDL-3.2 nuclear data library. The calculated β_eff value overestimated the experiments; the difference slightly exceeded the experimental error.     

Application of minimum plutonium criteria at EOC to optimize the fuel loading pattern in VVER-1000 reactors

Most of the strategies yet implemented to optimal fuel loading pattern design in nuclear power reactors, are based on maximizing the core effective multiplication factor (K_eff) to extract maximum energy and lowering the local power peaking factor (P_q) from a predetermined value. However, a new optimization criterion could be of interest, aiming at maximum burn-up of the plutonium content in fuel assemblies, i.e., minimization of remaining plutonium in spent fuel at the end of cycle (EOC). In this research, we developed a new strategy for optimal fuel core loading pattern of a VVER-1000 reactor, based on multi-objective optimization: lowering the P_q, maximization of the K_eff and minimization of remaining plutonium (Pu) in fuels at EOC. This strategy has been implemented considering exact calculations of fuel burn-up during the equilibrium cycle using WIMSD and CITATION calculation codes. We used the genetic algorithm to find the optimum fuel loading pattern. Simulation results show that this strategy can reduce the remaining Pu of the fuels at EOC while considering limitations on core power peaking and multiplication factor.     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback – a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies.     

Experimental study of neutronic parameters in Tehran research reactor mixed-core

In this work, general characteristics of a typical mixed core, including HEU & LEU fuel is studied. The study is performed in the Tehran research reactor (TRR). In this study the neutronic parameters, reactivity feedback coefficients and kinetic parameters are investigated. The reference core designated for such study is the equilibrium core (No. 61) with an average bun-up of 27% & 36% for SFE's & CFE's, respectively. The MTR_PC package is used for neutronic analysis. In this research, experimental and computational results for the reference and mixed core are compared. Meantime, the obtained values for neutronic parameters are mostly below the adopted safety criteria and they are in good agreement with the experimental results. However β_eff and ℓ_p are a little bit higher in the mixed core with respect to the reference core, but in practice, these small changes will not cause substantial impacts on the dynamic behaviour of the reactor core. The absolute values of the fuel temperature, moderator density and void coefficients of reactivity, are less in the mixed core and only the moderator temperature coefficient is higher. The calculated values of power defect, based on the reactivity coefficients; in both core configurations are in good agreement with the experimental values.