Equilibrium Cycle Research Articles

Analysis of actinide recycling in a Boiling Water Reactor using CASMO-4/SIMULATE-3

Final disposition of spent nuclear fuel in a permanent repository is still on debate; meanwhile the stockpile of depleted fuel is increasing. Actinide recycling has the potential to reduce that stockpile. This study analyzes the potential to burn actinide from depleted fuel in a thermal reactor without reducing the nominal electricity generation of the reactor and meeting all the safety constraints. The study takes an equilibrium cycle of a Boiling Water Reactor (BWR) as the reference cycle and then it proposes the design of an equivalent fuel assembly in which 4 fuel rods are replaced by minor actinide-bearing rods, the process ends when the equilibrium cycle is met. The new BWR fuel assembly design, that is called MA-c, has the goal to maximize the actinide recycling while introducing the lowest perturbation possible to the control and safety systems of the reactor. MA-c is successful, according to our analysis. The performance of the core is analyzed using MA-c and it is compared against the UO2 reference core, results show that both of these cores produce the same amount of energy for an 18-month cycle. The MA-c fuel assembly has a higher enrichment than the UO2 reference fuel assembly to compensate for the use of four MA-bearing rods, however all the fuel pins inside the MA-c fuel assembly have enrichments below 5 w/o of U-235. The MA-c core met all the safety BWR constraints without modifying any of the control systems. Results for a BWR with a nominal power of 2027 MWt show a reduction of 135.30 kg HM of minor actinide per 18-month operating cycle when the whole core is composed of “once-through” MA-c fuel assemblies. The use of this new BWR fuel assembly design can be used effectively for the reduction of minor actinide without the need of control systems modification in the reactor to meet safety limits and without affecting the energy production.

Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) with PBO Reflector

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2 /PuO2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U3 Si2 ). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I2 S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2 S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2 S-LWR design are U3 Si2 and advanced stainless steel respectively. In addition, the I2 S-LWR design adopts an integral configuration and a fully passive decay heat removal system to provide indefinite cooling capability for a class of accidents. This paper presents the equilibrium cycle core design and reactor physics behaviour of the I2 S-LWR with U3 Si2 and the advanced steel cladding. The results were obtained using the traditional two-stage approach, in which homogenized macroscopic cross-section sets were generated by WIMS and applied in a full 3D core solution with PANTHER. The results obtained with WIMS/PANTHER were compared against the Monte Carlo Serpent code developed by VTT and previously reported results for the I2 S-LWR. The results were found to be in a good agreement (e.g. <200 pcm in reactivity) among the compared codes, giving confidence that the WIMS/PANTHER reactor physics package can be reliably used in modelling advanced LWR systems.

Comparison of the PHISICS/RELAP5-3D Ring and Block Model Results for Phase I of the OECD/NEA MHTGR-350 Benchmark

The Parallel and Highly Innovative Simulation for INL Code System (PHISICS) has been under development at Idaho National Laboratory since 2010. It consists of several modules providing improved coupled core simulation capability: INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) (three-dimensional nodal transport core calculations); MRTAU (Multi-Reactor Transmutation Analysis Utility) (depletion and decay heat generation); and modules performing criticality searches, fuel shuffling, and generalized perturbation. Coupling of the PHISICS code suite to the thermal-hydraulic system code RELAP5-3D was finalized in 2013, and as part of the verification and validation effort, the first phase of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MHTGR-350 benchmark has now been completed.The theoretical basis and latest development status of the coupled PHISICS/RELAP5-3D tool are described in more detail in a concurrent paper. This paper provides an overview of the OECD/NEA MHTGR-350 benchmark and presents the results of exercises 2 and 3 defined for phase I. Exercise 2 required the modeling of a stand-alone thermal fluids solution at the end of equilibrium cycle for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The RELAP5-3D results of four subcases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 required a coupled neutronics and thermal fluids solution, and the PHISICS/RELAP5-3D code suite was used to calculate the results of two subcases.The main focus of this paper is a comparison of results obtained with the traditional RELAP5-3D “ring” model approach against a much more detailed model that includes kinetics feedback on individual “block” level and thermal feedbacks on a triangular submesh. The higher fidelity that can be obtained by this block model is illustrated with comparison results on the temperature, power density, and flux distributions. It is shown that the ring model leads to significantly lower fuel temperatures (up to 10%) when compared with the higher-fidelity block model and that the additional model development and run-time efforts are worth the gains obtained in the improved spatial temperature and flux distributions.

Studies on pH-dependent color variation and decomposition mechanism of Brilliant Green dye in Fenton reaction

Degradation of Brilliant Green (BG) dye has been studied using Fenton reagent in the dark environment. The intensity of dye color was pH dependent with maximum absorbance at pH 6. N-de-alkylation caused dye decomposition which resulted in the reduction of absorbance. The rate of BG dye decolorization increased with an increase in the H2O2 and Fe2+ concentration. The presence of background ions commonly found in an industrial dyeing wastewater suppressed the efficiency of color, COD and TOC removal within 9 %. There was about 2.9-fold increase in biodegradability index after 30 min of reaction. The proposed mechanism showed the routes of dye degradation with low errors (−0.88 to 0.49 g mol−1) with respect to the exact mass. Dye degradation began with N-de-ethylation, and later was further cleaved to aromatic-amine, -acid, -alcohol and -additive products. Hydroquinone was originated by hydroxylation of N,N-diethyl aniline. An equilibrium cycle among quinone, hydroquinone and Fe(III)-hydroquinone was exhibited through the formation of charge transfer complex.

TRIPOLI-4 Simulation of Core Physics Parameters for Two 3600-MW(thermal) Sodium-Cooled Fast Reactors

To support the development of Sodium-cooled Fast Reactors (SFRs) of Generation IV nuclear energy systems and to study the use of the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library on the core neutronics of large fast neutron reactors, in this work two recent Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) computational benchmarks of two 3600-MW(thermal) SFRs were analyzed with the continuous-energy TRIPOLI-4 code. Both a mixed oxide [(U,Pu)O2] core and a carbide [(U,Pu)C] core were investigated. Under two different fast neutron spectra, the reactor physics parameters—Keff, βeff (effective delayed neutron fraction), sodium void worth, Doppler constant, control rod worth, and core power distribution—were calculated for the beginning of equilibrium cycle condition. Both the pin-by-pin heterogeneous and fuel assembly–level homogeneous calculation models were applied in the whole-core simulation in order to evaluate their impact on the calculation results of SFR reactor physics parameters. The ENDF/B-VII.0 data library from the evaluation was also used with TRIPOLI-4 to study its impact on the SFR core reactivity and the boron carbide control rod worth. Using the mesh tally option, the energy deposition tally, and the upgraded display tool of TRIPOLI-4, radial power distribution and core power maps of the two cores were calculated and compared.

Single pass core design for a Super Fast Reactor

Single pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward coolant flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern significantly simplifies the core concept in terms of upper core structure, coolant flow scheme as well as refueling procedure. In single pass core design, supercritical-pressure water at approximately 25.0MPa enters the core at 280°C and heated up in one through upward flow to the average outlet temperature at 500°C. Great coolant density change in vertical direction will cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO2 also introduces great power increase in blanket assemblies during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of blanket assemblies to mitigate the power change in blanket assembly and to increase the coolant outlet. With exclusive refueling and shuffling scheme as well as partial length control rod design, neutronics & TH coupled calculation shows satisfactory results that can fulfil the requirement of design for both 500°C outlet temperature and negative coolant void reactivity.

Tolled multi-class traffic equilibria and toll sensitivities

We review properties of tolled equilibria in road networks, with users differing in their time values, and study corresponding sensitivities of equilibrium link flows w.r.t. tolls. Possible applications include modeling of individual travellers that have different trip purposes (e.g. work, business, leisure, etc.) and therefore perceive the relation between travel time and monetary cost in dissimilar ways. The typical objective is to reduce the total value of travel time (TVT) over all users. For first best congestion pricing, where all links in the network can be tolled, the solution can be internalized through marginal social cost (MSC) pricing. The MSC equilibrium typically has to be implemented through fixed tolls. The MSC as well as the fixed-toll equilibrium problems can be stated as optimization problems, which in general are convex in the fixed-toll case and non-convex in the MSC case. Thus, there may be several MSC equilibria. Second-best congestion pricing, where one only tolls a subset of the links, is much more complex, and equilibrium flows, times and TVT are not in general differentiable w.r.t. tolls in sub-routes used by several classes. For generic tolls, where the sets of shortest paths are stable, we show how to compute Jacobians (w.r.t positive tolls) of link flows and times as well as of the TVT. This can be used in descent schemes to find tolls that minimize the TVT at least locally. We further show that a condition of independent equilibrium cycles, together with a natural extension of the single class regularity condition of strict complementarity, leads to genericity, and hence existence of said Jacobians.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing Fast Reactors

We developed a simulation tool that accelerates the evaluation of design changes on the equilibrium cycle of fast-spectrum nuclear reactors. Within the tool, an implicit equilibrium cycle search is accelerated by a modal expansion perturbation method that expands arbitrary flux perturbations on a large basis of λ-eigenmode harmonics. The harmonics are computed only at the reference state using Krylov subspace iterative methods, and substantial perturbations from this state are shown to be well approximated by computationally efficient algebraic expressions. The modal expansion method is coupled to the equilibrium method to produce the later-in-time response of each design perturbation, resulting in an explicit perturbation-accelerated equilibrium cycle method.Because the method determines the perturbed flux explicitly, a wide variety of core performance metrics may be tracked within optimization frameworks, including the performance of thermal hydraulics, fuel, economics, core mechanical, and transients. This capability strongly differentiates the method from traditional generalized perturbation theory approaches. The motivating end-use of the method is to evaluate objective functions in multidisciplinary optimization of advanced reactor designs, though many other applications are envisioned.

Analysis of Loss-of-Flow Accidents for the NIST Research Reactor with Fuel Conversion from HEU to LEU

A program is underway to convert the current high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel in the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology. A RELAP5 model has been developed to analyze postulated accidents in the NBSR with the present HEU fuel and a proposed LEU fuel. The model includes the reactor vessel, primary pumps, shutdown pumps, various valves, heat exchangers, and average and hottest fuel elements and flow channels in the region where flow enters through an inner plenum (6 fuel elements) and a region where flow enters through an outer plenum (24 elements). The equilibrium cycle power distributions in the fuel elements were determined based on three-dimensional Monte Carlo neutron transport calculations performed with the MCNPX code. In this paper we discuss safety analyses conducted for the loss-of-flow accidents resulting from either loss of electrical power or inadvertent throttling of flow control valves at the inlets to the inner and outer plena. The analysis shows that the fuel conversion will not lead to significant changes in the safety analysis and that the calculated maximum clad temperatures, minimum critical heat flux ratios, and minimum onset of flow instability ratios assure that there is adequate margin to fuel failure.

A novel research reactor concept based on coated particle fuel

This work presents a preliminary study of a novel plate-type fuel concept for a high-performance and ultra-safe research reactor. This new fuel type consists of coated particle fuel (CPF) randomly dispersed in an aluminum matrix with a certain packing fraction that can be adjusted depending on the reactor design requirements. The CPF can also be varied in the fuel kernel material between UC and UO2. For the purpose of this study, UO2 was used as the reference fuel type. Using this novel fuel type, a 20MWth pool-type research reactor was investigated to determine the preliminary performance and safety characteristics of the new fuel. The core thermal analysis was done using the MATRA-P code. The neutronics analysis was done using the Monte Carlo Serpent code for an equilibrium cycle resulting from a multi-batch fuel management. In this analysis, it was found that the Doppler effect is significantly enhanced through the implementation of CPF, in turn improving the inherent safety of the reactor. In addition to the notable improvement in safety, the new fuel type also promises to be able to achieve a high thermal neutron flux, improving the performance and utility of the reactor. It is concluded that the CPF-based fuel concept presented in this paper can enable new high-flux reactor designs using simple plate-type fuels with improved safety.

HPLWR fine mesh core analysis

Abstract The European version of Supercritical Water Reactors (SCWR), the High Performance Light Water Reactor (HPLWR) operates in the thermodynamically supercritical region of water. Our basic objective was to elaborate a stationary coupled neutronic-thermohydraulic code capable for the calculation of the actual 3-pass core design with fuel assembly clusters. The calculations covered the neutronic transport calculations of HPLWR fuel assemblies, the coupled neutronic-thermohydraulic global calculations and the pin-wise analysis. Applying conservative assumptions, the relation to the linear heat rate and maximum cladding temperature limits was checked for the equilibrium cycle of HPLWR with this new code system.

Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

On the Use of Reduced-Moderation LWRs for Transuranic Isotope Burning in Thorium Fuel—II: Core Analysis

Multiple recycle of transuranic (TRU) isotopes in thermal reactors results in a degradation of the plutonium (Pu) fissile quality with buildup of higher actinides (e.g., Am, Cm, Cf), some of which are thermal absorbers. These phenomena lead to increasing amounts of Pu feed being required to sustain criticality and accordingly larger TRU content in the multirecycled fuel inventory, ultimately resulting in a positive moderator temperature coefficient (MTC) and void reactivity coefficient (VC). Because of the favorable impact fostered by use of thorium (Th) on these coefficients, the feasibility of Th-TRU multiple recycle in reduced-moderation (RM) pressurized water reactors (PWRs) and RM boiling water reactors (called RMPWRs and RBWRs, respectively) has been investigated. In this paper, Part II of two companion papers, the results of the single-assembly analyses of Part I are developed to investigate full-core feasibility. A large reduction in moderation is necessary to allow full actinide recycle. This increases the core pressure drop, which poses some thermal-hydraulic challenges, which are more pronounced if the design implementation is through retrofitting an existing PWR. For a given reactor cooling pump, the core flow rate is reduced. Despite this, it is possible to achieve feasible inlet and outlet temperatures and minimum departure from nucleate boiling ratio, for the reduction in moderation considered here. Reflood after loss-of-coolant accident is expected to be slower, which may lead to unacceptable peak clad temperatures and/or clad oxidation. Equilibrium cycles are presented for the RMPWR and RBWR, with a negative MTC and VC. However, the RMPWR may have positive reactivity when fully voided, and the hard spectrum makes it difficult to achieve an adequate shutdown margin, such that for the considered fuel designs, additional rod banks would be required.

The biomechanical properties of canine skin measured in situ by uniaxial extension

IntroductionA uniaxial extension system was setup to analyze the mechanical properties of dog skin. Material and methodsPads were glued onto dog skin with constant reproducible geometrical parameters and the extension force was measured as a function of the extension values. Forty-one sites (82 cycling tests) were investigated in situ on 11 canine cadavers, half of them after surgically isolating the test area from the surrounding skin. Series of loading-unloading cycles of up to 5N or 10N or both loads were performed on each site. The elastic properties and the dissipative effects were characterized respectively by evaluating the secant Rigidity at maximum loads and the Fraction of dissipated energy. ResultsA hysteresis phenomenon, implying the need for preconditioning to attain equilibrium cycles, was apparent during mechanical characterization. Polynomial expressions were used to relate the measured Rigidities and the Fractions of dissipated energy with or without sample isolation. The latter were less affected by isolation. The ratios between the Rigidities at 5N to those at 10N displayed non-linearity in the investigated extension range in contrary to the Fractions of dissipated energy. Discussion/conclusionThe parameters confirming the dissipative non-linear elastic behavior of dog skin were identified and the correlation between Rigidity and Fraction of dissipated energy on isolated and non-isolated skin samples was quantitatively determined. This extension setup can now be used as a “true in vivo” mapping tool to determine the mechanical characteristics of the skin during healing processes or during the study of Human skin disease with the dog as an animal model.

Studies on LLFP transmutation in a pressurized water reactor

A systematic study on the long-lived fission product (LLFP) transmutation in a pressurized water reactor (PWR) is performed, aiming at an optimal transmutation strategy for present nuclear energy development. The LLFPs selected in the analysis include 99Tc and 129I discharged from light water reactors (LWRs). The isotope 127I is also considered to avoid the difficulties in isotopes separation. To minimize the negative impacts of LLFPs on the core performance and safety parameters, metallic technetium or MgI2 target pins mixed with ZrH2 are designed and investigated. Through the numerical analysis on equilibrium cycles, the transmuted amounts of 99Tc and 129I equal to the yields from 1.94 and 4.22 PWRs with a power of 1000 MWe, respectively. Numerical results indicate that both 99Tc and 129I can be transmuted conveniently in present PWRs in the form of target pins.

Physics study of deep-burning of spent fuel transuranics using commercial LWR cores

This paper describes a physics study of the utilization and destruction of the transuranic (TRU) nuclides contained in spent fuel through a high burnup irradiation (Deep Burn) in commercial light water reactors (LWR). The TRU nuclides are contained in the Fully Ceramic Micro-encapsulated (FCM) fuel, comprising of TRISO particles dispersed in a matrix of silicon carbide and manufactured in fuel pellets and fuel pins. This study is performed by using the traditional two-step core design procedure, which consists of fuel assembly calculations and core analysis. The main focus is on the core physics characteristics including safety-related parameters. The target reference core is the YongGwang unit 3, cycle 6 reload core. New fuel assemblies consisting of the conventional UO2 pins and new FCM TRU fuel pins are charged into the cores from cycle 6 to an equilibrium cycle. The results show that the reload cores can be designed to achieve the deep-burning (∼60% burnup) of TRU without adverse changes to the safety-related parameters and the core performance parameters. The equilibrium core has TRU self-recycling capabilities, being able to “deep burn” as much TRU nuclides as it generates.

Information Stickiness in General Equilibrium and Endogenous Cycles

Abstract Traditionally, observed fluctuations in aggregate economic time series have been mainly modeled as being the result of exogenous disturbances. A better understanding of macroeconomic phenomena, however, surely requires looking directly at the relations between variables that may trigger endogenous nonlinearities. Several attempts to justify endogenous business cycles have appeared in the literature in the last few years, involving many types of different settings. This paper intends to contribute to such literature by investigating how we can modify the well-known information stickiness macro model, through the introduction of a couple of reasonable new assumptions, in order to trigger the emergence of endogenous fluctuations.

Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.