Linear Heat Generation Rate Research Articles

High-fidelity models for the online control of power ramps in fuel displacement systems – Revisiting the ISABELLE experiment in the OSIRIS Material Testing Reactor

The Jules Horowitz Reactor (JHR) is an MTR under construction at the CEA Cadarache, France. Its design and future operation build upon the lessons learned from the OSIRIS MTR. The CEA intends to transfer the knowledge accumulated with the ISABELLE1 loop of OSIRIS to the ADELINE loop of the JHR, both dedicated to power-ramp-type irradiation experiments, the purpose of which is to test the resistance of PWR fuel rod cladding under extreme levels of stress. In this article, we revisit the ETALISA experiment, a heat balance measurement experiment performed in 1992 for calibrating power ramps in ISABELLE1. We use high-fidelity modelling and simulation tools, especially the neutron-gamma TRIPOLI-4® Monte Carlo transport code, to calculate the detailed components of the heat balance, correction terms, and uncertainties. Comparisons between the simulations and the experiments show a very good agreement in the total linear heat generation rate of 400 W/cm at high power. The computed 2σ uncertainty is found to be 5%, a value essentially identical to the estimate derived in 1993 from an engineering approach. The use of modern simulation tools does not make it possible to improve upon this value, but provides a better understanding of the various components and corrections introduced in the total heat balance. The main limitations come from the ISABELLE1 online instrumentation, thermocouples and self-power neutron detectors, which set a limit on our very knowledge of the actual power ramp experimental conditions.

Extended Development of a Fission Gas Release Behavior Model Inside Spherical Fuel Grains for LWR Reactors

Fission gas plays a significant role in fuel rod performance following accidents. The amount of fission gas increases dramatically under accidental conditions. This leads to a subsequent rise in the fuel rod internal pressure and temperature due to aggravated gap conductance between the fuel pellet and cladding. As a result, fuel rod performance degrades. Therefore, studying fission gas behavior is crucial for accident assessment and evaluating fuel rod performance. Minimizing the Impact of fission gas on fuel rods is essential for maintaining their integrity and safety within nuclear reactors. One important aspect of ensuring safety is predicting fission gas release (FGR). In this study, we presented an extended model to be used in light water reactors (LWRs). The FGR can be modeled using COMSOL Multiphysics with the finite element method. This modeling approach considers both normal and abnormal conditions, with the latter categorized as Class-II type incidents. The model assumes that the gas diffusion inside a spherical grain varies over time. By examining perfect sinks with gas production, perfect sinks without gas production, and imperfect sinks under steady-state conditions, different initial and boundary conditions are set. To validate the accuracy and universality of expressions used in the model, input parameters from other models and experiments are utilized. By comparing the model’s results with these inputs, the accuracy and applicability of the expressions can be confirmed. This validation process ensures that the model provides reliable predictions for fission gas behavior in fuel rods under both normal and abnormal operating conditions. Based on our findings, it is evident that the FGR fraction displays an upward trend as diffusion coefficients and temperatures rise. Conversely, larger grain sizes and higher linear heat generation rates are associated with a reduction in the FGR fraction. Notably, enhanced resolution leads to a postponed onset of FGR. Furthermore, the influence of the diffusion coefficient on the FGR fraction primarily stems from the interconnected effects of temperature and linear heat generation rate.

Modeling of the P2M Past Fuel Melting Experiments with the Fuel Performance Code CYRANO3

CYRANO3 is the thermal-mechanical industrial code developed and used by Electricité de France (EDF) to simulate nuclear fuel rod performance under normal and transient conditions (power ramp tests) in pressurized water reactors and during transport and storage periods, as well. This code has already been successfully used by EDF for the last 30 years to justify normal operations and category 2 transients, covering various types of fuels: UO2, UO2 + gadolinium, mixed oxide, and various claddings, as well those proposed by nuclear fuel suppliers: Zircaloy-4, Zirlo,™ and Optimized Zirlo.™ The CYRANO3 code was recently improved to allow for modeling fuel melting. In this paper, a global overview of CYRANO3’s ability to simulate past power-to-melt (P2M) ramps is presented with a focus on recent developments carried out to assess fuel rod behavior under these conditions. CYRANO3 is demonstrated to be a powerful tool to provide reliable values of melted radii. As part of validation of these development works, CYRANO3 calculations have been used to assess two P2M ramp experiments carried out in the BR2 experimental core in Belgium (HBC-4 P2M ramp), and in the R2 experimental core in Sweden (xM3 P2M ramp). The main objectives of the work are to expand knowledge of the thermal-mechanical behavior of high-burnup fuel under P2M ramps by making interpretations of test simulations and to validate newly developed computational models for fuel melting that have been implemented in an extended version of the CYRANO3 fuel code. For both rods, the steady-state irradiation power history was captured and modeling was performed. The key results of the steady-state irradiation modeling are reproduced with fair accuracy by means of CYRANO3 simulations. The results demonstrate the good ability of CYRANO3 to simulate P2M ramps. The melted radii and conditions of failure are well predicted. Calculated melted radius at ramp terminal Linear Heat Generation Rate (LHGR) is in good agreement with the experimental measurements performed after the experiments.

Modeling of OECD/NEA P2M Benchmark Cases by Means of TRANSURANUS Code

This paper presents the fuel performance simulation of two past experiments that are concerned with partial fuel melting. The activity is conducted in the framework of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Power to Melt and Maneuverability (P2M) – Second Framework for Irradiation Experiments (FIDES-II) Joint ExpErimental Programme (JEEP). The xM3 transient experiment involved uranium oxide (UO2) fuel with a ZIRconium Low Oxidation alloy (ZIRLO) cladding base irradiated up to 27 MWd/kg U, followed a ramp test up to 70 kW/m leading to inception of melting. The High Burnup Chemistry Experiment 4 (HBC4) transient was performed according to a rapid power ramp, on a UO2 fuel with a Zircaloy-4 cladding base irradiated up to 47 MWd/kg U. The linear heat generation rate reached 66.3 kW/m at the end of the transient leading to inception of melting and rodlet failure. Three main cases per each of the two transients were modeled with the TRANSURANUS code version 2015 in agreement with the benchmark specifications to consider uncertainty on the power during the ramping phases. The paper presents and discusses the analysis of these tests and provides an in-depth sensitivity analysis to assess the capabilities of the code as well as the effect of user choices in simulating the base irradiation and the inception of melting in light water reactor fuel rods.

Fuel loading pattern optimization of ALLEGRO fast reactor using the tabu search metaheuristic technique

In this work an improved tabu search (TS) optimization technique is used to find the optimal fuel loading pattern in the experimental fast reactor ALLEGRO. The penalty function method is used to solve this multi-objective combinatorial optimization problem. The objective function and constrains are computed via an interface between the developed optimizer and the ERANOS 2.3 deterministic code. To reduce the running time, the operating cycle simulation is carried out using the diffusion approximation and a 7-group neutron energy structure. The proposed objective function maximizes the k-eff value at the end of cycle, satisfying the maximum power factor, the excess hot reactivity, and the linear heat generation rate constraints. A novel procedure developed in ERANOS to perform fuel assembly swaps for neighbor search is presented. The algorithm performed as expected throughout the iterative process and, unlike other metaheuristics based on local search, did not get stuck in local optima, which is a well-known advantage of tabu search technique.

Fuel loading pattern optimization of ALLEGRO fast reactor using genetic algorithms

In this work the genetic algorithmic (GA) optimization technique is used to find the optimal fuel loading pattern in the experimental fast reactor ALLEGRO. The penalty function method is used to solve this multi-objective combinatorial optimization problem. To evaluate each loading pattern, an interface between the genetic algorithm code and the ERANOS 2.3 deterministic neutronic simulation code was developed. For the simulation of the fuel cycle in ERANOS, the diffusion approach and a 7-group energy structure are used to reduce the computational cost. An integer variable encoding scheme was defined to represent the fuel loading patterns. Two crossover operators, the partially mapped crossover (PMX) and the order crossover (OX), combined with the random mutation operator were implemented as an alternative to the classical operators used in GA. To evaluate the performance of the code the results are compared using two objective functions and two crossover operators. In the first case the operating cycle length is maximized by maximizing the k-eff value at the EOC by satisfying the constraints on the power peaking factor over the cycle, the excess reactivity at the BOC, and the linear heat generation rate constraints. In addition, a condition was imposed that there should be no more than two contiguous positions in the pattern to preserve the power distribution uniformity and to avoid high power peaking factors. The code works very well, and a good solution was found in a reasonable computation time. The best solution corresponds to the PMX operator, and the running time is twice as short compared to the OX operator. In the second case study, a power distribution flatness term is included in the objective function. In this case, a more uniform radial power distribution is obtained but associated with a decrease in cycle length due to the penalty of the flatness term.

Adaptive second-order sliding-mode control for a pressurized water nuclear reactor in load following operation with Xenon oscillation suppression

This study proposes an adaptive second-order sliding mode control based on a twisting algorithm to control nuclear reactor core power during load-following power maneuvers. The control system was designed based on the concept of an extended equivalent control. The control technique does not require knowledge of the uncertainty or upper bound of the disturbance in the system. Additionally, the gain of the control system is a dynamic gain that increases or decreases according to the system requirements within a specified period. Consequently, chattering was attenuated. Chattering excites high-frequency dynamics and causes wear out of the control mechanism, making chattering a severe challenge in sliding-mode control. A nuclear reactor is a time-varying, complex, nonlinear, and constrained system. The characteristics of a nuclear reactor are a function of its operating power level, fuel burnup, and aging. In addition, the load-following mode of operation causes Xenon oscillation, which can further cause instability in the core. Therefore, the non-base load operation of the system, including load following, further aggravates uncertainty and disturbances in the system. To this end, an adaptive second-order sliding mode control that does not require knowledge of the uncertainty or the upper bound of the disturbance in the system was designed. The reactor core was modelled using an experimentally verified and validated multi-point kinetic model with four nodes. Simulation experiments were conducted using a multi-point kinetics model and an adaptive second-order sliding mode control. The results of the simulation experiments indicated that reactor core integrity is guaranteed, and the core is protected against peak power densities, such as linear heat generation rates (LHGRs) and lower departure from nucleate boiling ratios (DNBR). Moreover, the control system achieved the load-following objective and suppressed Xenon oscillation. The performance of the proposed control system was further compared with that of a twisting control system and classical proportional integral derivative control system to validate the effectiveness and reliability of the adaptive twisting second-order sliding mode control system.

BISON automation to predict transient power history based off prescribed cladding temperature

Since the restart of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) in 2017, advanced testing using various modules within the minimal activation retrievable capsule holder system has been underway with other types being considered for future work due to customer interest. In this work, one such test module—the Temperature Heat-sink Overpower Response module (THOR)—was simulated within BISON with UO2 fuel with zircaloy cladding and U-10Zr fuel with HT9 cladding, and a steel-alloy heat sink with a desired customer prescribed maximum cladding temperature profile. The THOR module is an instrumented capsule within which fuel rods can be irradiated in a liquid–sodium-filled hole in a thick-wall metal heat sink so that, along with TREAT’s unique transient power-shaping capability, experimenters can simulate numerous transient irradiation conditions. For transient testing, specimen temperature history is a primary driver in experiment design, which is primarily driven by the balance of power input to the nuclear heated specimen and the heat-transfer boundary condition associated with a particular experiment design. A simple iterative method was implemented in Python to write a BISON linear heat generation rate (LHGR) power history to produce desired customer-prescribed cladding-temperature values. The BISON LHGR power profile will later be used to write the control rod positions within TREAT to yield the customer-prescribed temperature values. Not all customer prescribed temperature values were achieved when desired temperatures were decreasing due to system limitations of THOR. However, the iterative method shown allows for temporal power histories to be derived, such that all BISON-produced temperatures are within 5% of customer values when increasing or maintaining a constant cladding temperature. The iterative method used allows for the dimensions, materials, and material properties of the THOR module to be customer specific, as well as different customer-prescribed parameters, such as fuel temperature, to be used in place of maximum cladding temperature. This tool allows for different designs and transient scenarios to be explored without manual fitting, allowing for rapid development for customer needs.

Numerical investigation of the thermal conductivity of UO2 – Mo microplate fuel pellets to realize enhanced heat transfer in the fuel radial direction

• Thermal conductivity of radially arranged UO 2 –3 vol% Mo microplate pellets increased. • Effective thermal conductivity increased by 53% at 1000 °C. • Central pellet temperature decreased by 149 °C under LHGR of 200 W/cm. • Recommend arranging Mo microplates parallel to the main heat transfer flow. The thermal performance of composite UO 2 fuel pellets, as a potential candidate for accident-tolerant fuels, is being attempted to be actively enhanced by employing high conductivity materials as additives. Herein, we numerically investigated the thermal performance of UO 2 – 3 vol% Mo microplate fuel pellets with microsized Mo plates to enhance the corresponding thermal conductivity in the fuel radial direction. UO 2 – 3 vol% Mo microplate fuel pellets were successfully fabricated through the conventional sintering process, and the characteristics of the thermal conductivities were investigated in terms of the shape factor of the Mo microplate, amount of Mo content, and arrangement (such as the spacing and angle) of the Mo microplates in the UO 2 fuel pellets. The results demonstrated that the arrangement of the Mo microplates parallel to the main heat transfer flow direction could further enhance the thermal conductivity. The numerical results pertaining to the calculation based on the microplate arrangement exhibited a reasonable agreement with the measured values, and the thermal conductivity was noted to be enhanced by 47% at 1000 °C compared to that of UO 2 . Moreover, the UO 2 – 3 vol% Mo microplate fuel pellets with enhanced thermal conductivities could reduce the maximum pellet temperature by 149 °C compared to that of the UO 2 pellet under a linear heat generation rate of 200 W/cm.

A methodology to enhance thermal neutron flux in Tehran Research Reactor core for domestic fuel test purposes

Performance of newly fabricated nuclear fuels must be tested under desired neutronic and thermal-hydraulic operating conditions prior to utilization in the reactor core. Tehran research reactor (TRR) as an MTR-type reactor is the only operating research reactor in the country that can potentially be used for irradiation tests on domestic fuels. In this study, compacting TRR core configuration, as a method to increase thermal neutron flux in the central in-core irradiation position to provide desired neutronic environment for fuel irradiation tests, is of concern. Several design criteria and limitations, including preservation of enough and suitable irradiation positions in the core for routine radioisotopes production activities of the reactor, have been considered to develop a new compact core configuration for TRR in which, desired linear heat generation rate is achieved in the test fuels and all the neutronic and thermal-hydraulic safety parameters of the proposed configuration are in accordance with the acceptance criteria stated in the reactor FSAR.

RECENT VALIDATION STUDY OF THE TECHNIQUE FOR EXPRESS EVALUATION OF BURNUP IN LEAKING FUEL ASSEMBLIES OF WWER POWER UNITS

The present work continues the series of papers on the revision of the conventional technique for evaluation of leaking fuel burnup during reactor operation at nuclear power plants (NPPs). The focus was made on reduction of uncertainties in evaluation of leaking fuel burnup in modern fuel cycles at WWER-1000 power units. A set of models was proposed for express calculation of the build-up of caesium isotopes in fuel and to relate 134Cs/137Cs activity ratio with fuel burnup for each rod in the core. These models are based on routine neutronic calculations of pin-by-pin linear heat generation rates which are performed at NPPs for each particular fuel cycle with particular core loading pattern (however, these calculations do not provide data on caesium inventory in fuel). Previously, the proposed models have been validated against several practical cases. This latest validation study relied on the analysis of the most recent fuel cycles at two NPPs that reported spike-events and identified the leaking fuel assemblies (LFAs) after the reactor shutdown. The calculated 134Cs/137Cs activity ratios in the fuel of the LFAs were compared to the NPPs data on the activity measurements, and to the post-irradiation examination (PIE) data provided for one FA. A reasonable agreement between the model predictions and the experimental data on 134Cs/137Cs activity ratios in the fuel as a function of its burnup is shown for the advanced FA designs in modern fuel cycles.

Nuclear data uncertainties for core parameters based on Swiss BWR operated cycles

The effects of the nuclear data uncertainties are quantified for specific core parameters from a real Boiling Water Reactor operated over 25 consecutive cycles. Nominal calculations are performed with the CASMO5 and SIMULATE-3 codes and the ENDF/B-VII.0 nuclear data libraries. The uncertainties are calculated using a Monte Carlo propagation method, and the ENDF/B-VII.1 covariance matrices. The conclusions are that the nuclear data uncertainties (1) have a limited impact on core parameters (axial power, linear heat generation rate, and comparison with in-core measurements), but (2) are larger for keff compared to the expectations from reactor experience. Additionally, the importance of simultaneously varying all input data in order to accurately calculate uncertainties is demonstrated.

Research on Improved Simplified SCWR Assembly and Core Design

Abstract A new type of simplified assembly without “water rod” or solid moderator is identified to simplify China Supercritical Water-cooled Reactor with the rated electric power of 1000 MWel (CSR1000) core structure. A study on the properties of this new assembly has been carried out. Under the condition of heat full power, maximum cladding surface temperature (MCST) is 658 °C, very close to safety criterion 650 °C, and maximum linear heat generation rate (MLHGR) is 40.0 kW/m, slightly higher than safety criterion 39 kW/m. Considering rod stuck criterion, the maximum keff is 0.9833, lower than safety criterion 0.99. The discharge burn-up reduced to 21368 MWd/t, a sharp decreasing of 31.1%. This simplified assembly provides a new idea for solving the problem of complex structure of supercritical water-cooled reactor (SCWR).

Conceptual design of Super FR for MA transmutation with axially heterogeneous core

Supercritical water cooled Fast Reactor (Super FR) is featured with the large coolant density reduction by almost 1/10 from the core inlet to the outlet. Since Minor Actinides (MAs) not only have large neutron capture cross sections for thermal neutrons, but also can fission with fast neutrons, MA transmutation performance of Super FR may greatly depend on MA loading positions in the core axial direction of Super FR.However, such investigations have never been conducted. Hence, this study aims to design a Super FR transmutation core concept with the axial configuration of multiple layers of Mixed Oxide (MOX) and blanket fuels, with a focus on the influence of the large axial coolant density change on MA transmutation and core characteristics.With the design criteria of the negative Void Reactivity Coefficient (VRC), the Maximum Cladding Surface Temperature (MCST) < 650 ℃ and the Maximum Linear Heat Generation Rate (MLHGR) < 39 kW/m, three-dimensional neutronics and thermal-hydraulics coupled core burnup calculations have been carried out. Assembly-wise coolant flow rate distribution is determined to attain high core average outlet temperature and the core characteristics of different designs have been evaluated for the equilibrium core after the cores have reached equilibrium states with given fuel shuffling schemes. It has been shown that the MA transmutation amount is limited by deterioration of VRC due to increase of Pu enrichment for compensating the reactivity penalty by MA loading. Moreover, such influence has been found to be more significant in the lower region of the core, where the coolant density is relatively high. Hence, the core design with MA loading to the upper MOX layer is favorable for improving the MA transmutation performance. However, the trade-off relationship between the MA transmutation amount and thermal-hydraulics performance (increase of MLHGR and decrease of average outlet temperature) has been revealed. To overcome the issue, the core radial zoning has been applied and it has been found effective to suppress the trade-off relationship.

VVER Thermal Power Measured Using the Readings of Rhodium Self-powered Neutron Detector (SPND)

A procedure for determining the linear heat generation rate in a fuel assembly of VVER reactor at the location of a rhodium self-powered neutron detector (SPND) and verification of this procedure are discussed. A method of measurements of the VVER thermal power based on SPND readings is described.

A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

The Advanced Test Reactor (ATR) has been used successfully for the testing of fast reactor fuel for nearly two decades. These successes have been in spite of numerous challenges for testing fast reactor fuel in the ATR (a thermal spectrum reactor), but the solutions to those challenges have resulted in excessively long irradiation times (~10 years) for high-burnup targets as well as experiments that are highly sensitive to fabrication tolerances and eccentricities. This paper presents a solution to the problems of extended irradiation times and fabrication sensitivities. Thermal and neutronic analyses were performed to show that a reduced-diameter fuel pin with an equivalent linear heat generation rate can provide a prototypic thermal profile (peak centerline and inner clad temperature) along with a near-prototypic power profile within the ATR thermal spectrum. This allows the experiment to reach a high burnup in an expeditious timeframe compared to traditional ATR fast fuel irradiations. In addition, problems with fabrication sensitivities were addressed by introducing a double-encapsulated experiment that pushes the high heat flux helium gap farther away from the fuel pin. Fuel pin position eccentricities are also mitigated by using a large sodium bond between the pin and capsule fuel. The advantages and potential pitfalls of this revised design are discussed, including the effect of length scales on fuel system behavior.

Numerical and experimental investigation on thermal expansion of UO2-5 vol% Mo microcell pellet for qualitative comparison to UO2 pellet

Thermally conductive microcell UO2 pellets in which metallic networks enclose UO2 granules have been highlighted due to their enhanced thermal performance for a potential candidate of accident-tolerant fuels. However, studies on thermal expansion effects of microcell UO2 pellets are still lacking. Here, stress and strain distributions of the fabricated UO2-5 vol% Mo microcell pellet under thermal expansion were investigated. After we confirmed that the numerical results agreed well with experimental measurement in terms of linear thermal expansion, stress and strain distributions of the microcell pellets were calculated under maximum stress and maximum strain conditions. As a result, formation of Mo network decreases the maximum hoop stress and the maximum radial strain of UO2 pellets by 50.9% and by 29.4%, respectively under a linear heat generation rate of 200 W/cm. Finally, the UO2-5 vol% Mo microcell pellet can decrease the maximum cladding stresses by 36.7%.

Механизмы и главные причины повреждений тепловыделяющих элементов и тепловыделяющих сборок атомных электрических станций с реакторами с водой под давлением

This article is the first part of the review of publications on the problem of increasing the service life and improving the reliability of fuel assemblies (FA) and fuel rods (FR). In the next issues of the journal, Part 2 Dynamic interactions between the coolant flow and FAs and Part 3 Prediction of vibroacoustic resonances in the reactor cores of VVER-based nuclear power of this review will be published. This review part presents the results from investigations into the causes and mechanisms of FR failures, in which radioactive materials (gaseous fission products and volatile elements, in particular, krypton, xenon, iodine, and cesium) escape from the fuel pellet into the reactor primary coolant circuit. It is shown that the nuclear fuel depressurization level is influenced by all stages of the product life cycle, including design, manufacture, and operation of FAs. In particular, during operation, FA vibrations have an effect on the nuclear fuel depressurization level. As is well-known, an FA is a hydroelastic oscillatory system consisting of two interacting subsystems: a mechanical one and a hydrodynamic one. The FA vibration is due to the dynamic processes occurring in these subsystems. It is noted that the hydrodynamic subsystem has been studied to a rather insufficient extent, which is due to complexity of describing the processes through which the random hydrodynamic loads applied to the streamlined surfaces are produced, as well as the influence of flow thermal-hydraulic and acoustic characteristics on the occurrence of self-excited oscillations. Based on the defective FA operation data, the service life of leaky fuel rods and their linear heat generation rates are estimated. The evolutions of fuel operation conditions and fuel design changes are described. The available recommendations on optimizing FA designs developed taking into account the experience gained from operation of the reactors and the results of the accomplished fretting wear investigations are mentioned. It is shown that all who manufacture and operate nuclear fuel take efforts aimed at improving its quality, because it is the key indicator characterizing the vendor of fuel and nuclear power plants in the world market. Achieving longer fuel campaigns and higher burnup values while keeping or even improving the safety level are the most important criteria for nuclear fuel quality. Currently, the project called Nuclear Fuel Zero Failure Level is in the stage of implementation and obtaining the first results of joint activities of all participants.

Effect of Dual Surface Cooling on the Temperature Distribution of a Nuclear Fuel Pellet

Heat removal from nuclear reactor core has been one of the major Engineering considerations in the construction of nuclear power plant. At the center of this consideration is the nuclear fuel pellet whose burning efficiency determines the rate of heat transfer to the coolant. This research, focuses on the study of temperature distribution of solid fuel, temperature distribution of annular fuel with external cooling and the temperature distribution of annular fuel with internal and external cooling. We analyzed the different distribution and made a conclusion on the possibility of improving temperature management of Nuclear fuel rod, by designing fuel pellets based on this geometrical and thermal Analysis. To date, a lot of studies has been done on the thermal and geometrical properties of Nuclear fuel pellet, it is observed that annular fuel pellet with simulteneous internal and external cooling can achieve better temperature distribution which leads to high linear heat generation rate, thus generating more power in the design [1]. It has also been observed that annular fuel pellets has low fission gas release [10]. In large LOCA, the peak cladding temperature of annular fuel is about 600 which is significantly less than that of solid fuel (920 ), this is due to the fact that annular fuel cladding has lower initial temperature and the thinner annular fuel can be cooled more efficiently than the solid fuel. One of drawbacks of annular fuel technology is “the fuel gap conductance assymmetry” which is caused by outward thermal expansion, it has a potential effect on the MDNBR (Minimum Departure from Nucleate Boiling Ratio), which is the minimum ratio of the critical to actual heat flux found in the core [10]. In this model, we used the ceramic fuel pellet of UO2 as our case study. All the parameters in this model are assumed parameters of UO2. The Heat Transfer tool (ANSYS APDL) was used to validate the Analytical Model of this research.

Improved Core Design of a High Breeding Fast Reactor Cooled by Supercritical Pressure Light Water

The authors look for an attractive light water reactor (LWR) concept, which achieves high breeding performance with respect to the compound system doubling time (CSDT). In the preceding study, a high breeding fast reactor concept, cooled by supercritical pressure light water (Super FBR), was developed using tightly packed fuel assembly (TPFA) concept, in which fuel rods were arranged in a hexagonal lattice and packed by contacting each other. However, the designed concept had characteristics, which had to be improved, such as low power density (7.4 kW/m), large core pressure loss (1.02 MPa), low discharge burnup (core average: 8 GWd/t), and low coolant temperature rise in the core (38 °C). The aim of this study is to clarify the main issues associated with improvement of the Super FBR with respect to these design parameters and to show the improved design. The core design is carried out by fully coupled three-dimensional neutronics and single-channel thermal-hydraulic core calculations. The design criteria are negative void reactivity, maximum linear heat generation rate (MLHGR) of 39 kW/m, and maximum cladding surface temperature (MCST) of 650 °C for advanced stainless steel. The results show that significant improvement is possible with respect to the core thermal-hydraulic characteristics with minimal deterioration of CSDT by replacing TPFA with the commonly acknowledged hexagonal tight lattice fuel assembly (TLFA). Further design studies are necessary to improve the core enthalpy rise by reducing the radial power swing and power peaking.