Current Pressurized Water Reactors Research Articles

Heat Transfer Analysis of Nanofluid Based Coolant Used in the Sub-Channel of Fuel Assembly of a Pressurized Water Reactor

Nano fluids are found as one of the important suspension nanoparticles in the solution that show a very significant improvement on (boiling) critical heat flux (CHF) at moderate concentrations of nanoparticles. CHF is considerate to be the maximum limit of nucleate boiling. Moreover, CHF is the most essential factor for improving the heat transfer mode, and thus the reactor performance. The goal of this study is to investigate the use of nanofluid as a component of the primary coolant in the pressurized water reactor (PWR) to improve heat transfer. This enhances the heat transfer characteristics of the reactor core and also prevents the core from melting in an emergency situation. In current and future PWRs, the nano fluid application could allow substantial power upgrades, thereby enhancing their economic efficiency. In addition, the use of nanofluids could allow as much as 20 percent increase in power density in latest generation plants without any change in the design of the fuel assembly. In this analysis, it has been found that almost 1–4% increase of Nano particle with base fluid caused a substantial escalation in heat transfer, which can reduce the overall bulk temperature and the temperature of the fuel rods as well. In addition to this, turbulent kinetic energy and velocity have been developed and investigated for different percentages of Nano fluid along the sub channel of a PWR fuel assembly, which is crucial in case of design. Overall, there has not been done much CFD based work regarding the Nano fluid application in the coolant of the sub channel, thus impact of nano fluid inclusion in the base water for heat transfer escalation inside a sub channel of a PWR is a crucial topic for investigation, in addition to this, necessary Computational Fluid Dynamics (CFD) relevant data base has been generated for further investigation. Hence, the present CFD analysis represents the novelty and uniqueness of the work for the inclusion of nanofluid with water.

Long term comparison between reprocessed nuclear fuel cycles

Based on the idea of adopting closed fuel cycle in current pressurized water reactors (PWR) in order to reduce the use of natural uranium and recycle the spent fuel accumulated in the world inventory, this paper aims to compare two closed nuclear fuel cycles simulated at Model for Energy Supply Strategy Alternatives and their General Environmental Impacts (MESSAGE). The nuclear fuel cycles compared are: i) a closed fuel cycle with recovering of plutonium (Pu) to fabricate the mixed oxide (MOX) fuel; ii) a closed fuel cycle with recovering of a transuranic matrix to fabricate the transuranic fuel spiked with depleted uranium (TRU-U)O2. The comparison is based on the Brazilian nuclear energy system. They consider the time frame of 2019-2060 and the introduction of Angra 3 in the system. Advantages and disadvantages of using the strategy of operating with the different nuclear fuel cycles are shown, which include results regarding natural uranium consumption, spent fuel accumulation or utilization, nuclear waste and the nuclear fuel costs for both fuels.

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.

On the validation of BEPU methodologies through the simulation of integral experiments: Application to the PKL test facility

Validation is a key step in the field of thermal hydraulics and has played a crucial role in the implementation of system codes. These codes are then included in licensing methodologies being those either best estimate plus uncertainties (BEPU) or conservative approaches. However, only a small amount of efforts has been done on the validation of the methodologies as a whole. Such validation would imply the comparison of the bounding limits and the experimental data. Due to the lack of actual data of accidental situations, such validation should be carried out with data from integral test facility (ITF) experiments.The OECD/NEA PKL-4 test programme is investigating safety issues relevant for current pressurized water reactor (PWR) plants as well as for new PWR design concepts focusing on complex heat transfer mechanisms under two-phase flow. Within the project, a blind benchmark activity has been launched where an IBLOCA experiment has been simulated by 12 different organizations. The Universitat Politècnica de Catalunya (UPC) participated in the activity with a blind BEPU approach with the main objective of assessing the validity of the methodology as a whole.At first, the scalars that characterize the phenomenology of the particular scenario have been considered. Secondly, in order to broaden the extent of the validation, the results of the different participants have been interpreted as if they were blind experiments as well. Finally, different output parameters have been considered providing a matrix of results that is used to evaluate the validity of the methodology.

Preliminary study on physical characteristics of single‐pass super‐critical water‐cooled reactor core

The super-critical water-cooled reactor (SCWR) is one of six types of Gen-IV reactors. Based on the current mature pressurized water reactor technology, it has several advantages, including high conversion ratio, high breeding of nuclear fuel, high thermal efficiency, and a simplified system. However, due to the combination of water coolant and fast neutron spectrum, there is a need to address safety problems caused by the positive void coefficient. The purpose of this paper is to give a study on physical characteristics for the reactor core. Herein, the core modeling of an SCWR with a single-pass flow is that seed assemblies and blanket assemblies are arranged in a tornado manner, in which MCNP and SERPENT are used for critical calculation. The assembly design is discussed from the aspects of the thickness of the solid moderator layer and cladding material. Based on the optimized assembly design, UO2, PuO2-ThO2, and PuO2-UO2 are selected for the sensitivity analysis of fuel composition. In addition, sensitivity analysis for axial coolant density distribution, water-to-fuel volume ratio, reflector thickness and fuel arrangement are carried out. Several fundamental system characteristics are discussed in this paper, including keff, void coefficient, Doppler coefficient, neutron spectrum, power distribution, one-cycle burnup and conversion ratio. The results help in understanding the design of the SCWR better and provide fundamental research data for further study.

Utilization of 3D fuel modeling capability of BISON to derive new insights in performance of advanced PWR fuel concepts

The Internally and eXternally cooled Annular Fuel (IXAF) and Lightbridge's metallic Helical Cruciform Fuel (HCF) are two innovative fuels, which could allow significant power uprate (∼20%) in current Pressurized Water Reactors (PWRs) while maintaining or improving safety margins. Both fuels would substantially benefit from 3D modeling to properly assess their performance in a PWR core and to further their technology readiness level. For IXAF, the possible misalignment of the annular pellets creates a mismatch in inner and outer gap size, which will result in an azimuthally asymmetric temperature distribution and different heat split between the inner and outer channel. For HCF, the fuel deformation requires 3D consideration due to its inherit helical geometry. In this work, the first attempt of building thermo-mechanical capability for analysis of these fuels was made in BISON fuel performance code. The 3D fuel behavior under normal operation along with loss of coolant accident (LOCA) and reactivity initiated accident (RIA) were simulated. The preliminary key insights derived from the 3D fuel performance modeling were the potential excessive ballooning of IXAF during LOCA and excessive swelling of HCF during RIA. However, overall both fuels did show promising and improved performance at 20% uprated conditions compared to traditional solid fuel geometry and material.

Numerical Simulation Research on the Performance of SCWR Fuel Rod

Because of the high temperature and high pressure characteristics of supercritical water-cooled reactor (SCWR), the thermal hydraulic performance of SCWR is greatly different from pressurized water reactor (PWR), which makes the current PWR fuel rod performance analysis codes are no longer applicable to SCWR. In this research, the irradiation swelling, irradiation densification, thermal expansion, thermal creep, plastic deformation, irradiation creep and irradiation hardening of UO2 pellet, and stainless steel cladding were considered; the gas conductance and radiant conductance of gap heat transfer were considered, the forced convective heat transfer on the outer surface of cladding was considered. Meanwhile, the irradiation effects and the thermal effects on the materials parameters such as thermal conductivity, specific heat, and young’s modulus were also considered in this research. With the help of abaqus software, the related user-defined subroutines were developed, and the irradiation effects and thermal effects of SCWR fuel were introduced into the numerical simulation, and then completed the analysis of SCWR fuel rods’ performance under steady power conditions. Some reference suggestions for the design and development of SCWR fuel could be provided by the establishment of this numerical simulation method.

Fundamentals of reactor physics with a view to the (possible) futures of nuclear energy

Abstract This paper present basic nuclear reactor physics that may help to understand next challenges that nuclear industry have to face in the future. The ones considered in this work are waste production and natural resources consumption. This paper shows that waste and resources are linked by the plutonium status that could be considered as the principal waste or a valuable material that should be saved for a future transition to breeder reactors that could be for instance Sodium-cooled Fast Reactors (SFRs). This kind of reactors does not rely on natural resource supply, as it produces its own fissile material, plutonium-239, after a neutron capture on uranium-238. However, the operation of SFRs needs a huge amount of plutonium that should be produced in current Pressurized Water Reactors (PWRs). Natural uranium available on earth is expected to allow the operation of the global current fleet until approximately 2100 without major issues. The transition from PWRs to SFRs is then needed if and only if the nuclear industry will face an increase at a global scale during this century. If not, plutonium, the most radiotoxic element produced in PWRs, should be considered as a waste. Consequently, the plutonium status depends on the future evolution of nuclear energy. This paper shows that a status-quo means that plutonium should be considered as the most problematic element that should be handled on a long-term basis. On the other hand, a strong increase in nuclear energy on a global scale would imply that plutonium is a valuable material that would make a transition to SFRs possible.

The Powerful Method of Characteristics Module in Advanced Neutronics Lattice Code KYLIN-2

Due to powerful geometry treatment capability, method of characteristics (MOC) currently becomes one of the best method to solve neutron transport equation. In MOC method, boundary condition treatment, complex geometry representation, and advanced acceleration method are the key techniques to develop a powerful MOC code to solve complex problem. In this paper, we developed a powerful MOC module, which can treat different boundary conditions with two methods. For problems with special border shapes and boundary condition, such as rectangle, 1/8 of square, hexagon, 1/3 of hexagon, 1/6 of hexagon problems with reflection, rotation, and translation boundary condition, the MOC module adopts periodic tracking method, with which rays can return to start point after a certain distance. For problems with general border shapes, the MOC module uses ray prolongation method, which can treat arbitrary border shapes and boundary conditions. Meanwhile, graphic user interface based on computer aided design (CAD) software is developed to generate the geometry description file, in which geometry is represented by “lines and arcs” method. With the graphic user interface, the geometry and mesh can be described and modified correctly and fast. In order to accelerate the MOC transport calculation, the generalized coarse mesh finite difference (GCMFD) is used, which can use irregular coarse mesh diffusion method to accelerate the transport equation. The MOC module was incorporated into advanced neutronics lattice code KYLIN-2, which was developed by Nuclear Power Institute of China (NPIC) and used to simulate the assembly of current pressurized water reactor (PWR) and advanced reactors, to solve the transport equation with multigroup energy structure in cross sections database. The numerical results show that the KYLIN-2 code can be used to calculate 2D neutron transport problems in reactor accurately and fast. In future, the KYLIN-2 code will be released and gradually become the main neutron transport lattice code in NPIC.

Proposal for improved nuclear fuel utilisation and economic performance by utilising thorium

A systematic and strategic nuclear power reactor deployment roadmap has been developed for South Africa within the national strategic plan, utilizing thorium-based fuel. The roadmap was developed through analysis of economical, strategic and historical aspects. The accumulated advantages of thorium-based fuels are summarized, which could form the initiative to implement thorium-based nuclear fuels in South Africa. A timeline (which forms the basis of the roadmap) was constructed and consists of three different phases. Phase 1 starts in 2015 and extends to 2030. Phase 2 starts in 2031 and ends in 2044 whilst Phase 3 is from 2045 to 2060. Each phase is discussed with regard to construction, implementation and research activities. This roadmap starts at current pressurized water reactors (PWRs) and advances to future reactor technologies, using an evolutionary approach. In addition to the results reported in this paper, the economic advantages to introducing thorium as a fertile component in PWR fuels as compared to once-through conventional uranium-only cycles is explored (Du Toit & Cilliers, 2014). The economic evaluation compares uranium fuel to thorium-uranium fuel in terms of the fuel cycle costs, reactor downtime costs due to refuelling and income derived from electricity sales.

Assembly Design of Pressurized Water Reactors with Fully Ceramic Microencapsulated Fuel

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design-basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly.This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCM fuel in a 17 × 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to-diameter ratio (P/D) and downselected two alternate assembly designs: 11 × 11 (1.65-cm outer diameter and 1.18 P/D) and 9 × 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.

Power maximization method for land-transportable fully passive lead–bismuth cooled small modular reactor systems

Although current pressurized water reactors (PWRs) have significantly contributed to global energy supply, PWR technology has not been considered a trustworthy energy solution owing to its problems of spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, a lead–bismuth eutectic (LBE) fully passive cooling small modular reactor (SMR) system is suggested. This technology can not only provide the solution for the problems of SNFs through the transmutation feature of the LBE coolant, but also strengthen safety and economy through the concept of natural circulation cooling SMRs. It is necessary to maximize the advantages, namely safety and economy, of this type of nuclear power plants for broader applications in the future. Accordingly, the objective of this study is to maximize the reactor core power while satisfying the limitations of shipping size, materials endurance, and criticality of a long-burning core as well as safety under beyond design basis events. To achieve these objectives, the design limitations of natural circulating LBE-cooling SMRs are derived. Then, the power maximization method is developed based on obtaining the design limitations. The results of this study are expected to contribute to the effectiveness of the reactor design stage by providing insights to designers, as well as by formulating methods for the power maximization of other types of SMRs.

A DFM/Fuzzy/ATHEANA Human Failure Analysis of a Digital Control System for a Pressurizer

A methodology comprising Dynamic Flowgraph Methodology (DFM) and A Technique for Human Error Analysis (ATHEANA) is applied to a digital control system proposed for the pressurizer of current pressurized water reactor plants. The methodology consists of modeling this control system and its interactions with the controlled process and operator through an integrated DFM/ATHEANA approach. The results were complemented by the opinions of experts in conjunction with fuzzy theory. In terms of human reliability, DFM, along with ATHEANA, can model equipment failure modes, operator errors (omission/commission), and human factors that, combined with plant conditions, influence human performance. The results show that the methodology provides an efficient fault analysis of digital systems identifying all possible interactions among components. Through prime implicants, the methodology shows the event combinations that lead to system failure. Quantitative results obtained are in agreement with literature data, with a few percentage value discrepancies.

Response of equal channel angular extrusion processed ultrafine-grained T91 steel subjected to high temperature heavy ion irradiation

The life extension of current pressurized water reactors and the design of reliable next-generation nuclear reactors call for advanced structural steels that can sustain radiation up to several hundred displacements per atom (dpa) at elevated temperatures. Here we performed Fe ion irradiation to 150dpa at 450°C on bulk coarse-grained (CG, with a grain size of ∼2μm) and ultrafine-grained (UFG, with grain size of ∼320nm) T91 steels. Extensive microscopy studies show that fine grains in UFG T91 reduced the density of nanocavities and dislocation loops. The swelling rate in UFG steel is three times lower than that of CG T91 due to the existence of abundant defect sinks, such as high angle grain boundaries and dislocations. A strong surface effect with size dependence was noted during heavy ion irradiation studies. The large deviation of swelling rate from neutron irradiated specimens implies the significance of He concentration and presumably dose rate on swelling in nuclear reactors.

Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: 235U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes.

Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor

SCWR (Supercritical Water Reactor) is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was used to calculate the thermodynamic properties of water and steam. The ISO-5167-4: 2003 standard was incorporated in the code as the basis of orifice plate to compute the flow rate. New heat balance model and uncertainty estimate have also been included in the code. In order to validate H-Power, an assessment was carried out by using data published by US and Qinshan Phase II. The results showed that H-Power was able to estimate the thermal power of SCWR.

Core Design, Spent-Fuel Characteristics Assessment, and Fuel Cycle Analysis for Thorium-Uranium Breeding Recycle in Pressurized Water Reactors

This work is focused on core design, spent-fuel characteristics assessment, and fuel cycle analysis for thorium-uranium breeding recycle in a typical pressurized water reactor (PWR), without any major change to the fuel lattice and the core internals but substituting the uranium oxide (UOX) pellet with a thorium-based fuel pellet. Two mixed cores are investigated, one loaded with mixed reactor-grade plutonium-thorium oxide (PuThOX) fuel assemblies and the other with mixed reactor-grade 233U-thorium oxide (U3ThOX) fuel assemblies. The high purity of reactor-grade 233U extracted from burnt PuThOX fuel is used as seeds of U3ThOX for starting thorium-uranium breeding recycle.The core design and analysis indicated that thorium-uranium breeding recycle is technically feasible in current PWRs. In the mixed core with U3ThOX loading, the well-designed U3ThOX assemblies were located on the periphery of the core as a “blanket” region, which remain in core for six cycles and get breeding with 232Th-233U. The feedback parameters and kinetic parameters are dominated by the UOX fuel in the inner core. For the UOX/PuThOX mixed core, the higher plutonium content leads to harder neutron spectrum, smaller reactivity worth of neutron absorbers, and smaller delayed neutron fraction and prompt neutron lifetime, which are similar to the current mixed cores partially loaded with the plutonium-uranium mixed-oxide (MOX) fuel.The fuel cycle analysis has shown that 233U monorecycling with U3ThOX fuel could save 13% of natural uranium resource compared with UOX once-through fuel cycle, slightly more than that of plutonium monorecycling with MOX fuel. If 233U multirecycling with U3ThOX fuel is implemented, more natural uranium resource would be saved.

Life cycle assessment of the European pressurized reactor and the influence of different fuel cycle strategies

This article presents a life cycle assessment (LCA) of electricity generated by the European pressurized reactor (EPR), a Generation III (Gen III) nuclear reactor concept. The focus of the LCA is on the potential locations in Switzerland and France in the year 2030. Environmental burdens from the EPR are also compared to those from current pressurized water reactor (PWR) technology in Switzerland. The entire fuel cycle is analysed and includes improved modelling of natural uranium mining. The indicators employed in the impact assessment characterize a broad range of burdens to the environment and human health at both the global and the regional/local level and includes an analysis on the basis of external costs. The LCA shows a reduction of greenhouse gas emissions by around just 7 per cent for the EPR in Switzerland in 2030 compared to the current PWR cycle emitting just above 5 g CO2-eq/kWh. Due to differences in cooling strategies, which cause slightly different output capacities, the French EPR is expected to achieve marginally better reductions of almost 16 per cent. The processing of natural uranium to yellowcake was shown to be the stage in the nuclear fuel cycle having the highest burdens and potential impacts across a majority of indicators, accounting for between 25 per cent and 60 per cent of burdens and potential impacts. External costs of normal operation are shown to be around just 2 per cent of estimated internal (production) costs and similar to those of hydro and offshore wind. The potential impacts due to nuclear-related accidents were not considered.

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants

The negative impact of a boron dilution accident on the safety of a current pressurized water reactor (PWR) initiated investigations with the aim of checking the feasibility of reduced boron concentration operation. In addition, reduction of the maximum boron concentration in a PWR is a practical and feasible means to substantially reduce the radiation dose to operators and to minimize corrosion damage. Four types of integral burnable absorbers have been considered: gadolinium, integral fuel burnable absorber (IFBA), erbia, and alumina boron carbide. Under consideration of four different kinds of fuel assemblies (FA), four core design candidates were developed by applying current PWR OPR-1000 technology and by keeping major engineering design constraints and the equivalent fuel enrichment level used in the reference core (REF) design. However, an optimal design was targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between the REF and the optimal core designs is presented here. One of the designs is suggested as the most promising and favorable low boron core (LBC) design in this framework. The proper combination of axial and radial enrichment zoning patterns plus a mixture of fresh FAs with depleted assemblies in an LBC design candidate with an IFBA-bearing FA at equilibrium cycle could bring a two times narrower axial offset variation than that of the REF design, maintain an acceptable power peaking factor ˜23% lower than the design limit, and achieve higher fuel burnup. It was observed that this optimal LBC design could comply with current OPR-1000 reactor acceptance criteria associated with smooth reactivity swing, more flattened power distribution, and desired limiting safety parameters despite an 18% loss of shutdown reactivity worth at beginning of cycle when compared to the REF design.

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs. The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.