Minimum Departure From Nucleate Boiling Ratio Research Articles

Transient full core analysis of PWR with multi-scale and multi-physics approach

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

Evaluation of DNBR with neutronics calculation in LWR systems

The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.

Subchannel thermal hydraulic analysis of 5 × 5 rod bundle with CRUD layer

This paper studies the thermal–hydraulic behaviors difference with or without the Chalk River Unidentified Deposits (CRUD) layer by using subchannel analysis code COBRA-EN. Based on similar structure characteristics, the EPRI CHF prediction model is improved and applied in a 5 × 5 rod bundle. Three different species transient scenarios are calculated for comparing different hydraulic behaviors with or without the CRUD layer. The results show that the heat transfer enhancement is limited under high pressure because the improved CHF model is mainly affected by local pressure. But when the pressure came low, the enhancement became more obvious, such as heat flux and minimum departure from nucleate boiling ratio (MDNBR). Under serious accidents, the influence can be mitigated with the CRUD layer, and recovery from the accident with the CRUD layer was faster than bare cladding when considering the effect of the CRUD layer.

Engineering uncertainty analysis of the heat transfer of fuel plates

Uranium-containing aluminum-cladding dispersion plate fuels are widely used in various research reactors owing to their excellent heat-transfer capability. In the nuclear design stage, both the maximum temperature of the cladding surface and the minimum departure from nucleate boiling ratio (MDNBR) must satisfy safety acceptance criteria to ensure the performance of the fuel plates. Many engineering-related uncertainties lead to deviations between the actual and design values. For safety considerations, these engineering-related uncertainties must be considered when analyzing the maximum temperature of the cladding surface and MDNBR. In this study, the various uncertainties of the engineering hot channel and hot spot factors, three methods of uncertainty analysis, and sensitivity analysis of these uncertainty factors were investigated. The results show that the square root of the sum of squares method is both conservative and economical. The sensitivity analysis results show that the flow channel gap deviation, produced by fuel plate manufacturing, has the greatest influence on the maximum temperature of the fuel cladding surface.

PREDICTION OF AP1000’S NUCLEAR REACTOR PRESSURE VESSEL TEMPERATURE DURING NORMAL OPERATION

Modeling of thermal-hydraulic calculations for the AP1000 core to predict the reactor pressure vessel (RPV) temperature has been carried out. The reactor’s primary coolant system transfers the heat produced in the reactor fuel during reactor operation to the steam generator. Part of the heat will also be transferred from the coolant to the reactor vessel and the pipe. This paper presents the calculation result of the RPV temperature prediction during AP1000 normal operation. Calculations were performed using COBRA-EN code for analyzing the core thermal hydraulics and using analytics for predicting the RPV temperature. These methods were carried out with the aim to predict the RPV temperature as well as at steady state nominal power conditions, at the function of flow, and at power fluctuation conditions. The calculation results at nominal power 3400 MWt (100% heat generated in fuel was assumed) and thermal design flow with 10% tube plugging (TDF2) of 48,443.7 ton/hr, for the minimum system pressure of 15.1 MPa, nominal system pressure of 15.513 MPa, and design system pressure of 17.133 MPa, show that the core outlet coolant temperature is 326.96°C, 327.01°C, and 327.22°C, and the RPV temperature is 303.65°C, 303.87°C, and 306.67°C, and the minimum departure from nucleate boiling ratio (MDNBR) is 3.21, 3.29, and 3.01, respectively. During reactor operation at a fixed nominal power of 3400 MWt, nominal system pressure, and under the condition of flow fluctuation, the maximum RPV temperature is shown to be 303.87°C.

Investigation of hybrid nanofluids effects on heat transfer characteristics in VVER-1000 nuclear reactor

In this study, the analysis of heat transfer characteristics of VVER-1000 nuclear reactor by using hybrid nanofluids as a coolants in subchannels in hot fuel assembly have been performed. Firstly, the criticality (keff) and fuel assembly power distributions of the VVER-1000 reactor by using Monte Carlo particle transport (MCNP) code to determine of the heat transfer characteristics with COBRA IV PC code have been examined. Secondly, the axial coolant temperature distributions and Departure from Nucleate Boiling Ratio (DNBR) values in hottest channel for Cu, CuO and TiO2 nanoparticles of the different volume fractions (0.05%, 0.01% and 0.1%) mixed with 0.1 vol % Al2O3 hybrid nanofluids have been analyzed. The obtained results by using hybrid nanofluids have been compared with the pure water results. The keff decreases with the increase of volumetric fractions of hybrid nanoparticles Cu, CuO and TiO2 mixed with 0.1% Al2O3 in the coolant. The maximum temperature has been reached by the püre water coolant at the end of the channel as 612.68 K whereas, the highest temperature change has been obtained in the case of 0.1% Al2O3 + 0.1% TiO2 as 617.35 K when evaluated among nanoparticle themselves. The results of analysis show that the heat transfer enhances with increasing hybrid nanoparticles volume fractions in the subchannel geometry. Also, the obtained minimum DNBR values using hybrid nanofluid haven't been exceed the safe operational limits; therefore the reactor operation would be safer.

Steady state thermal hydraulic modelling of WWR-S tank-in-pool research reactor for the purpose of its power upgrading

Abstract This paper presents a newly developed steady state core thermal hydraulic model (named SSTH-RR10 model) for upgrading the Egyptian first Research Reactor (ETRR-1), from its original power of 2 MWth to a higher level of 10 MWth, by considering different types of nuclear fuels. The SSTH-RR10 model is capable to predict and calculate, by means of a developed computer program, all the steady state thermal hydraulic parameters for the defined core configuration for each fuel type at 10 MWth. Three different fuel types were investigated: the reference fuel EK-10 rod type, the MTR plate type, and the IRT-4M ducted type. For each fuel type, the distribution of central fuel, clad, and coolant temperatures for average and hot channels of the core were predicted in the axial direction. Power distributions and pressure gradients were predicted as well. Moreover, the program calculates the safety limits and margins against the critical phenomena encountered such as the Onset of Nucleate Boiling (ONB), Departure from Nucleate Boiling (DNB), and the Onset of Flow Instability (OFI). Results of the SSTH-RR10 program for benchmarks of powers of 2 and 10 MWth are verified by comparing it with the published results of the International Atomic Energy Agency (IAEA), and those published for other programs such as PARET code, and very good agreement is found. The safety margins against ONB and DNB were evaluated in which the minimum DNB ratio was found to be about 3.1, which gives a sufficient margin against the DNB. The present work gives confidence in the model results and applications.

Searching for the most optimum burnable absorbers (BAs) for AP-1000 from the neutronic, thermal-hydraulic, and solid mechanics points of view

Because of the importance of absorber materials in regulating the reactivity of AP-1000, several materials with varied designs have been investigated to find the most optimum absorber materials from the neutronic, thermal-hydraulic, and solid mechanics perspectives. In this work, ZrB2, Er2O3, (SiO2-B2O3) and (Al2O3-B4C) were investigated as burnable absorbers (BAs) and H3BO3, Ag–In–Cd, and SS-304 as non-burnable absorbers. For the investigated BAs, the main neutronic parameters such as the fuel burn-up, axial thermal neutron flux distribution and the pin by pin distribution have been simulated. For the non-burnable absorbers, the control rod worth (CRW) and the boron concentration coefficient have been calculated. In the thermal-hydraulic analysis, the temperature distribution of the fuel, clad, coolant and the minimum departure from nucleate boiling ratio (MDNBR) have been studied. The BAs’ effects on solid mechanics analysis have emerged on both the von-Mises stress and the maximum fuel outer surface displacement.

Investigating the neutronic, thermal-hydraulic, and solid mechanics analysis for AP-1000 nuclear reactor

ABSTRACT This study investigates the neutronic, thermal-hydraulic, and solid mechanics analyses for one of the central fuel assemblies with 2.35% U235 enrichment in the AP-1000 Reactor. MCNPX 2.7 computer code, which depends on a Monte Carlo algorithm, was applied to perform the neutronic calculations. Both COMSOL-Multiphysics, based on the numerical solution for the centered rod sub-channel model, and MATLAB, based on the analytical solution for the same model, were applied to perform the thermal-hydraulic analysis. COMSOL-Multiphysics was used to perform the solid mechanics analysis. The neutronic, thermal hydraulics, and solid mechanics analyses play an essential role in the control, design, and safety of nuclear reactors. The calculations of criticality, reactivity, fuel burn-up, neutron flux, and power mapping were performed.‎ From the power mapping, the hot channel was determined and its maximum PPF. Upon getting the maximum PPF, which occurred in channel number 30 and equals 1.357, the thermal-hydraulic calculations using both COMSOL-Multiphysics and MATLAB were applied to determine the temperature distribution of the fuel, clad, and coolant. The departure from nucleate boiling ratio (DNBR) distribution was calculated to determine the minimum departure from ‎nucleate boiling ratio (MDNBR). The MDNBR results were 2.154 based on the analytical approach and 2.20 based on ‎the numerical approach. These values are higher than the limiting value of 1.75, which assure that no nucleate boiling ‎problems will occur. The solid mechanics calculations were investigated by coupling three kinds of physics using COMSOL-Multiphysics (the heat transfer in solid physics to perform the thermal load, turbulence flow RANS K-OMEGA to perform the pressure load, and solid mechanics physics). This coupling was used to determine the von-Mises stress, volumetric strain, and the total displacement for both the fuel and the clad material. The total fuel outer surface displacement was calculated to assure that no surface contact would happen between the ‎fuel and clad material to avoid chemical interactions between them. The displacement was calculated to be 0.007 mm, which is less than the He-gap ‎thickness, 0.008 mm. The coupling between different physics is one of the most well-developed techniques used for simulating the exact situation inside nuclear reactors. This coupling was done using the computational fluid dynamics (CFD) approach. All the calculations for the benchmark problem (a single fuel Assembly (G-9), near Beginning of Life, hot Full Power, equilibrium Xenon, and unrodded Core) were conducted. The obtained results were within the safety limits outlined in the design control document (DCD). Eventually, the results were in line with earlier efforts that utilized different codes for simulating neutronic analysis.

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Improved departure from nucleate boiling prediction in rod bundles using a physics-informed machine learning-aided framework

The critical heat flux (CHF) corresponding to the departure from nucleate boiling (DNB) crisis is a regulatory limit for the licensing of pressurized water reactors (PWRs) worldwide. Despite the abundance of predictive tools available to the reactor thermal-hydraulics community, the path for an accurate CHF model remains elusive. This work approaches the prediction of DNB through a physics-informed machine learning-aided framework (PIMLAF) with the objective of achieving superior predictive capabilities for a rod bundle. In view of the limitations in existing macro-scale physics-driven tools, an improved mechanistic model is first proposed, leveraging key concepts in the liquid sublayer dryout and bubble crowding mechanisms. The proposed mechanistic model is able to predict DNB in different heater geometries for a broad range of flow conditions without the need for recalibration. This model is then incorporated as the physics-informed component of the hybrid framework PIMLAF, which takes advantage of established understanding in the field (i.e., domain knowledge [DK]) and uses machine learning (ML) to capture undiscovered information from the mismatch between the actual and DK-predicted output. Two bundle-related case studies using the PWR subchannel and bundle tests (PSBT) database are carried out to illustrate the PIMLAF’s improved performance over traditional approaches for both interpolation and extrapolation purposes. In light of the PIMLAF’s promising potential to reduce prediction error, reactor vendors are encouraged to leverage their in-house experimental efforts and apply the hybrid framework to potentially achieve margin reductions in the minimum DNB ratio (MDNBR) for the designs of interest.

Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation

Abstract Since estimating the minimum departure from nucleate boiling ratio (MDNBR) requires complex calculations, an alternative method has always been considered. One of these methods is neural network. In this study, the Back Propagation Neural network (BPN) and Radial Basis Function Neural network (RBFN) are introduced and compared in order to estimate MDNBR of the VVER-1000 light water reactor. In these networks, the MDNBR were predicted with the inputs including core mass flux, core inlet temperature, pressure, reactor power level and position of the control rods. To obtain the data required to design these neural networks, an externally coupledcode was developed and its ability to estimate the thermo-hydraulic parameters of the VVER-1000 reactor was compared with other numerical solutions of this benchmark and the Final Safety Analysis Report (FSAR). After ensuring the accuracy of this coupled-code, MDNBR was calculated for 272 different conditions of reactor operating, and it was used to design BPN and RBFN. Comparison of these two neural networks revealed that when the output SMEs of the two systems were approximately the same, the training process in RBFN was much faster than in BPN and the maximum network error in RBFN was less than in BPN.

Thermal-hydraulic and solid mechanics safety analysis for VVER-1000 reactor using analytical and CFD approaches

Neutronic, thermal-hydraulics, and solid mechanics computations were used to study the different VVER-1000 reactors' operational conditions for the steady-state and the transient state. The main objective of these studies is to investigate the safe operating conditions, which is essential in determining the reactor safety limits. The coupling between different physics enables a better understanding of reactor phenomena and enhanced accuracy by reducing the assumption needed for separate calculations for different physics. MCNPX code is used to provide the source power distribution for two themes of calculations. The first theme is the analytical solution using MATLAB software, and the second theme is the finite element coupled physics using COMSOL Multiphysics. Analytical solutions using MATLAB software and results from COMSOL-Multiphysics were used to simulate the uncoupled analytical models and coupled models, respectively. The coupled theme provide feedbacks between different physics by iterating the solution between different physics; such as solving the heat transfer with taking into consideration the variation of the coolant velocity along the coolant channel and solving solid mechanics with taking into consideration the change of either the fuel and the clad temperatures distribution (thermal load) or the change in the coolant pressure (pressure load). While in the uncoupled solutions, certain assumptions reflect an average behavior for the effect of other physics on the physics under concern, such as the average pressure of the coolant and the average value for the coolant velocity. The obtained results are verified by comparing it to WIMS D-4 for neutronic calculations and COPERA-EN code for the thermal-hydraulic analytical solution. Also, thermal-hydraulic results by both coupled and uncoupled themes are verified against results from the final safety analysis report. The Departure from Nucleate Boiling Ratio (DNBR) and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) were applied for the investigation of VVER-1000 reactor safety parameters. Solid mechanics, fluid dynamics and heat transfer were used to compute the maximum stress/maximum volumetric strain acting on both fuel and cladding material. Results from both methods satisfied the limits addressed by the designer and shows a good agreement with the other published works that used WIMS D-4 for neutronic calculations and COBRA-EN code for thermal-hydraulic calculations analytically. in addition, the maximum stress occurring on the fuel and clad materials didn't exceed the yield stress. Furthermore, the fuel material displacement didn't exceed the Helium gap thickness, which means the surface contact between the fuel and the clad material is impossible that prevents the chemical interaction between both fuel and clad at 1135 K.

Preliminary neutronic/thermal-hydraulic evaluation and safety system optimization of PWR loaded with fully ceramic microencapsulated fuel

As one of the Accident Tolerant Fuels (ATFs), Fully Ceramic Microencapsulated (FCM) presents advantageous safety performance in the severe accidents compared with UO2 owing to its high melting point, superior heat transfer ability and proliferation resistance capability. In this study, preliminary neutronic and thermal-hydraulic analysis of Chinese improved three-loop Pressurized Water Reactor (PWR) with FCM pellet and SiC cladding (FCM-SiC) were conducted to demonstrate the fuel performance of ATFs. First, a three dimensional full core model loaded with FCM-SiC was developed to acquire the cycle length, power distribution and isotopic inventories. And low reactor power and high fuel enrichment were used to satisfy the annual reload requirement. Then, the reactor thermal-hydraulic safety evaluation was carried out by the modified subchannel and system analysis codes, which depicts lower Fuel Centerline Temperature (FCT) and Peak Cladding Temperature (PCT) in FCM-SiC reactor compared to the reactor with traditional fuels. The Minimum Departure from Nucleate Boiling Ratio (MDNBR) of cladding is much higher than the criterion of 1.3 as well. Furthermore, the reactor safety system was simplified and optimized by taking full advantage of the preponderance of ATFs, and the results show that the reactor of FCM-SiC can operate safely with only 10% initial flowrate of Low Pressure Safety Injection (LPSI) or without Accumulators (ACCs) during Large Break Loss of Coolant Accident (LBLOCA), while the conventional zircaloy is failed under the same conditions. This work could provide guidelines for the optimized design of the reactor safety system in the future.

Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation

Abstract Since estimating the minimum departure from nucleate boiling ratio (MDNBR) requires complex calculations, an alternative method has always been considered. One of these methods is neural network. In this study, the Back Propagation Neural network (BPN) and Radial Basis Function Neural network (RBFN) are introduced and compared in order to estimate MDNBR of the VVER-1000 light water reactor. In these networks, the MDNBR were predicted with the inputs including core mass flux, core inlet temperature, pressure, reactor power level and position of the control rods. To obtain the data required to design these neural networks, an externally coupledcode was developed and its ability to estimate the thermo-hydraulic parameters of the VVER-1000 reactor was compared with other numerical solutions of this benchmark and the Final Safety Analysis Report (FSAR). After ensuring the accuracy of this coupled-code, MDNBR was calculated for 272 different conditions of reactor operating, and it was used to design BPN and RBFN. Comparison of these two neural networks revealed that when the output SMEs of the two systems were approximately the same, the training process in RBFN was much faster than in BPN and the maximum network error in RBFN was less than in BPN.

Development and assessment of a new rod-bundle CHF correlation for China fuel assemblies

One of the main objectives of subchannel analysis for a pressurized water reactor (PWR) core or fuel assembly is to predict the critical heat flux (CHF) which usually employs the correlation based on the local parameters. The effects of spacer grid, cold wall, and non-uniform heat flux are essential for the sake of accurate calculation of local parameters. The subchannel code CORTH is applied to determine the local thermal-hydraulic parameters at the occurrence of departure from nucleate boiling (DNB) for each set of bundle CHF data of China fuel (CF) assemblies. In order to develop a new CHF correlation for CF assembly, the minimum DNB ratio (MDNBR) point and the burn-out (BO) point methods are both employed. The three-step form was adopted for the correlation which considers the effects of spacer grids, cold walls, and non-uniform heat flux. The coefficients of the correlation were determined based on the nonlinear regression method. The analysis and assessment results against the experimental data indicate that the correlation can predict the CHF with high prediction accuracy and correct parametric trends. In comparison with MDNBR point method, the BO point method is recommended to predict the risk of DNB because it is more reasonable and conservative and has a better predication rate. When implemented with the subchannel code CORTH, the developed correlation can predict the thermal-hydraulic performances of CF assemblies.

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

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

Core designing of the newly proposed (U+Gd)O2 FAs in the VVERs core and comparison with current UO2FAs

This research is an independent study on the Russian newly proposed FAs (Fuel Assembly) called TVS-2M, contains (U+Gd)O2 for the future advanced VVERs. The TVS-2M FAs contain gadolinium-oxide which is mixed with UO2 (by different Gd density and U-235 enrichments which are given herein), but also the core does not contain BARs (Burnable Absorbers Rods). The new type TVS-2M FAs are modeled by the SARCS software package to find the PMAXS format for three states of CZP (Cold Zero Power) and HZP (Hot Zero Power) as well as HFP (Hot Full Power), and then the whole core is simulated by the PARCS (Shiraz-university Advanced Regenerator of Cross Section library) code to investigate transient conditions. In addition, the WIMS-D5 code is suggested for steady core modeling includes TVS-2M FAs and/or current TVS FAs. Several neutronics aspects such as the first cycle length (first cycle burnup in terms of MWth-day/kg-U), the critical concentration of boric-acid at the BOC (Beginning of Cycle) as well as the axial and radial power peaking factors, differential and integral worth of the most reactive CPS-CRs (Control Protection System-Control Rod), reactivity coefficients of the fuel, moderator, and boric-acid, and the core’s under-moderation estimation are carried out, and also benchmarked with the PWR Core Design Basis Limits. Moreover, thermal-hydraulics core design criteria such as MDNBR (Minimum Departure from Nucleate Boiling Ratio) (based on W3 correlation), and the maximum of fuel and clad temperatures (radially and axially) are investigated herein. The results show that all of the mentioned parameters are in acceptable safety margins and in some aspects the TVS-2M design has superiority on TVS design such as more discharge burnup (more cycle burnup) and more intrinsic safety margins.

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.

Application of regression, variance, and density based global sensitivity methods to integrated VERA-CS and BISON simulations

Uncertainty quantification (UQ) and sensitivity analyses (SA) are performed for coupled simulations between VERA-CS, a coupled pin resolved neutron transport and subchannel thermal hydraulics code, and the fuel performance code BISON. The interface between VERA-CS and BISON is performed in a multiphysics environment known as the LOCA Toolkit for U.S. light water reactors (LOTUS) currently under development at Idaho National Laboratory (INL). A focus is placed on using a variety of SA measures, including two regression based (Pearson and Spearman), one variance based (Sobol indices), and three moment independent measures (Delta moment independent measures with L1, L2, and L∞ norms). The problem under inspection is a single assembly depletion case for three fuel cycles. The figures of merit are the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). SA results show MDNBR to be linear with consistent rankings throughout the fuel cycles. MFCT is linear, but with a change in rankings at the switch from open gap to closed gap models. GCPP is nonlinear at intermediate states that coincide with the onset of contact between fuel and cladding. These nonlinear states allow for the showcasing of higher order SA measures over first order methods.