Departure From Nucleate Boiling Ratio Research Articles

Safety analysis of pressurized water reactors with annular fuel during different transients

The dual-cooled annular fuel has two channels, with a shorter heat transfer path and a larger heat transfer area than solid fuel, which can effectively improve the safety of the reactor. It has excellent potential to improve the economy of the pressurized water reactor (PWR) and is considered to have a good development prospect. In this paper, the safety of the reactor with annular fuel assemblies under different transient conditions is analyzed using the system code NUSOL-SYS and the subchannel code NACAF, with the CPR1000 as the reference power plant. The analysis mainly includes the large break loss of coolant accident (LBLOCA), main steam line break accident (MSLB), loss of flow accident (LOFA), and rod ejection accident (REA) under nominal power and higher power density. The flow blockage accident and the combined accidents of LBLOCA and the flow blockage accident are also evaluated. The results show that during LBLOCA, the peak cladding temperature (PCT) of the core is only 833 K, much lower than that of the solid fuel. In addition, the departure from nucleate boiling ratio (DNBR) of annular fuel is larger during MSLB and LOFA, which means a more negligible risk of fuel damage. In REA, the average temperature of annular fuel is below 900 K, and the average enthalpy is below 200 kJ/kg, which is much lower than that of solid fuel. The results at 150 % power show that the performance of the core with annular fuel is still better than that of solid fuel at nominal power, except that the PCT during LBLOCA is slightly higher than solid fuel at nominal power. During the flow blockage accident, the results with different blockage areas show that a small proportion of blockage of the inner channel does not have a significant effect on the cladding temperature and that there is a risk of rupture when the inner channel is completely blocked. Finally, the results of the assumed combined accident show that the blockage of the inner channel before scram will significantly increase the PCT of the core and the time to complete the reflood during LBLOCA.

Thermal-Hydraulic modelling and analysis of VVER-1200 reactor core

A mathematical model has been developed to simulate the thermal–hydraulic performance of the VVER-1200 pressurized water reactor under normal operation. The energy equation and the heat conduction equation are solved analytically in order to predict the coolant, clad and fuel temperature distributions. The core active length is divided into axial regions and the fuel rod is divided into radial zones, nodal calculation is performed for three types of cooling channels; the core mean channel with mean core power, and both the mean and hot channels of the most intensive energy fuel assembly. Through this model, the heat flux leading to the Departure from Nucleate Boiling (DNB) as well as the Departure from Nucleate Boiling Ratio (DNBR) predicted at each axial node for each channel using EPRI correlation. High accuracy correlations and/or models valid under the reactor operating conditions are selected to estimate the heat transfer coefficients under single-phase forced convection, subcooled boiling and bulk boiling regimes. The vapor quality and void fraction are estimated for boiling regimes as well. The model is then used to simulate the VVER-1200 reactor performance under both the rated and conservative operational parameters where boiling is predicted in the hot channel under both cases. The results are verified against the available calculations in the literature for coolant temperature, clad temperature and void fraction, where a very good agreement is achieved. The developed model is considered an appropriate tool for independent thermal–hydraulic safety assessment of VVER-1200 reactor core under steady-state operation.

Thermal hydraulic preliminary calculations for small modular reactor based on ACPR50S technology

This paper presents preliminary thermal-hydraulics calculation results for a small modular reactor (SMR) with a thermal capacity of 200 MW based on the ACPR50S reactor technology (Advanced customer-friendly practicable reliable 50 MWe compact SMR). The basic parameters and thermal-hydraulics safety were assessed, including the heat coolant flow through the core, fuel, and cladding temperatures to ensure the reactor’s capability to operate within safe limits. The SMR configuration was simulated using the RELAP5 program and performed the calculation in about 500 seconds in the steady state. The coolant temperature of the heat carrier along the height of the core at the inner and outer was found to be different (1.7%) from the reference value. Regarding safety, the departure from nucleate boiling ratio (DNBR) was also estimated along the average channel and the hottest channel of the core in the steady state. The results from this study will be further used in transient computations and incident analysis for SMRs.

Design and optimization of dual-cooled fuel assembly in a 12×12 configuration for NuScale SMR based on neutronic-thermohydraulic parameters using the combined ANN-GA approach

The Small Modular Reactors (SMRs) use conventional solid fuels. To increase the safety of SMRs, Dual-Cooled Fuel (DCF) can be used. The DCF with two heat transfer surfaces improves heat removal from the fuel. For the first time, this study investigated the use of DCF in a new fuel assembly configuration for the NuScale reactor. Important neutronic and thermohydraulic (N–Th) parameters are obtained. Finally, the optimal DCF rod in a 12 × 12 configuration for the NuScale reactor has been presented. For this purpose, initially, a DCF assembly in a 12 × 12 configuration was designed for the reactor. The fuel mass, enrichment of 235U, and dimensions of fuel assemblies and the core are kept fixed. Then, 32 types of fuel rods with different internal radii were proposed for this assembly. Each of these fuels was placed separately in the assembly and simulated by neutronic codes under cold-clean and full power conditions. Also, the designed fuels were simulated using Computational fluid dynamics (CFD) software, and important thermohydraulic parameters were obtained. Also, to select the optimal fuel rod, the Artificial Neural Network (ANN) and Genetic Algorithm (GA) coupled method is used. To investigate the optimal DCF effects on N–Th parameters in the NuScale reactor, the reference reactor was also simulated and compared with the results of optimal DCF. As one of the most important results of this study, the fuel rod with an internal radius of 0.413423 cm was introduced as the optimal fuel. Also, optimal fuel reduces the temperature and pressure drop, and increases the Departure from Nucleate Boiling Ratio (DNBR).

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.

Optimization for fuel loading and reactivity initiated accident analysis for TRR-1/M1

TRR-1/M1 reactor core consists of two types of standard TRIGA fuel, 8.5% and 20% weight uranium. Periodically, the reactor core configuration is rearranged in order to optimize its performance. The objective of this work is to show that it is safe to operate TRR-1/M1 at 1 MW steady state condition for the new core configurations of which the 8.5% weight fuels in the B ring are replaced by high burn up 20% weight fuel elements. To ensure sufficient safety margins of the reactor operation, thermal–hydraulic computer codes, i.e., COOLOD-N2 and EUREKA-2/RR, were coupled with neutronic calculation based on Monte Carlo N-Particle Transport concept in order to analyse some crucial operating parameters, e.g., coolant and fuel temperature in the reactor core as well as Departure from Nucleate Boiling Ratio (DNBR) along the fuel axial direction of the hottest channel. From this study, it can be concluded that no significant effect on the reactor operation was found when replacing the 8.5% weight fuel elements in the B ring with the high burnup 20% weight fuels. Transient phenomena of Reactivity Initiated Accident (RIA) were also performed for the rapid removal of a sample from the reactor core with reactivity insertion rate of $1.00/s. It is found that the negative temperature coefficient of TRIGA fuel provided safety for the reactor even during the prescribed reactor transients. The calculation showed slightly increase in the fuel temperature during the short period of reactivity insertion which caused insignificant energy release to heat up the fuel during the accident. The maximum fuel temperature found during the accident in this study was well below TRR-1/M1 safety.

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.

Safety Analysis of 250-kW Philippine Research Reactor-1 Thermal-hydraulics under Steady-state Operations Using MARS-KS Code

Julius Federico M. Jecong1,3,4*, Alvie A. Astronomo2, Frederick C. Hila1, Neil Raymund D. Guillermo1, and Sweng Woong Woo4 1Applied Physics Research Section; 2Nuclear Reactor Operations Section Department of Science and Technology–Philippine Nuclear Research Institute Quezon City, Metro Manila 1101 Philippines 3Nuclear and Quantum Engineering Department Korea Advanced Institute of Science and Technology 291 Daehak-ro, Yuseong-gu, Daejeon 34141 Republic of Korea 4Reactor and Safety Evaluation Department Korea Institute of Nuclear Safety 62 Gwahak-ro, Yuseong-gu, Daejeon 34142, Republic of Korea Safety Analysis of 250-kW Philippine Research Reactor-1 Thermal-hydraulics under Steady-state Operations Using MARS-KS Code Keywords: 250-kW TRIGA reactor, MARS-KS, safety analysis, steady-state, thermal-hydraulics The Philippine Nuclear Research Institute (PNRI) of the Department of Science and Technology (DOST) is implementing a project to use the Philippine Research Reactor-1 (PRR-1) TRIGA nuclear fuel in a subcritical reactor (SCR) for training, education, and research in nuclear science and technology. However, although an SCR is a valuable training and research facility, it offers limited industrial applications. To potentially reap more benefits, DOST-PNRI is also investigating the feasibility of converting the zero-power SCR to a low power critical reactor in the future. In this work, the thermal-hydraulic behavior of a hypothetical 250-kW low power configuration of the PRR-1 TRIGA reactor is investigated using the MARS-KS code. At this power level, the fuel centerline or the hottest region of the fuel was determined to have 236.85- °C temperature during steady-state conditions. This temperature is well below the 749.85-°C temperature limit of TRIGA fuel that will result in excessive swelling. The peak heat flux occurs at the axial center of the fuel rod while the calculated departure from nucleate boiling ratio (DNBR) of 6.18 is minimum at this location. Moreover, the maximum fuel temperature is found to be insensitive to variations in coolant velocity, which ranges from 0.20–0.26 m/s within the natural circulation conditions of this facility. Based on the Bernath critical heat flux (CHF) correlation, coolant velocities below 0.2 m/s will still yield a DNBR above 5.0. These results demonstrate that a 250-kW PRR-1 TRIGA reactor operating under steady-state conditions is capable of safely removing the heat generated in the fuel

Neutronics and both analytical and numerical solutions for the rod centered subchannel thermal-hydraulic model

Abstract This work investigates the capacity to detect safety levels in pressurized water reactors using neutronic and thermal-hydraulic calculations (PWR). The main neutronic parameters, such as criticality calculations, reactivity calculations, burn-up calculations, and power distribution calculations, have been calculated by using MCNPX based on the Monte Carlo method. The hot channel was determined from the power distribution. On the hot channel, the rod-centered sub-channel model was used to investigate the main thermal-hydraulic parameters such as the fuel, cladding, and coolant temperature distribution in both axial and radial directions, the coolant pressure drop, and the departure from nucleate boiling ratio (DNBR). Two techniques were used to solve the rod-centered sub-channel model. MATLAB code was used for the analytical methodology, and COMSOL-Multiphysics computer software was used for the numerical methodology. The numerical technique has demonstrated high accuracy in determining the reactor safety level consistent with the analytical approach.

Subchannel analysis of thermal-hydraulic performance in rod bundle with spacer grid considering anisotropic turbulent mixing

Combined with the relationship between the time averaged parameters of flow field and the inter-subchannel transverse turbulent mixing velocity, a method for solving the mixing velocity between adjacent subchannels is presented in this paper. Then an anisotropic mixing model, which is based on subchannel-wise inhomogeneous turbulent mixing, is presented to reflect the three-dimensional mixing effect in rod bundle with spacer grid. By comparing the anisotropic mixing model with isotropic mixing model, it is found that the anisotropic mixing model agrees well with the experimental data from literature. In addition, based on the anisotropic mixing model, analysis on the subchannels with different spacer grids is also performed, including configurations with or without the mixing vanes and different arrangements of mixing vanes. It is found that the forced mixing effect between subchannels is mainly caused by the mixing vanes of the spacer grid. Besides, the continuous change of the flow and temperature fields is also caused by mixing vanes, which improves the Departure from Nucleate Boiling Ratio (DNBR) of the rod bundle channel. Also, the effect of different mixing vane arrangements on flow and heat transfer in the rod bundle is analyzed in detail.

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.

Assessment of Realistic Departure from Nucleate Boiling Ratio (DNBR) Considering Uncertainty Quantification of Core Flow Asymmetry

Concern over the asymmetric phenomena in the core region has increased considering safety issues that are highly possible to reduce the thermal margin significantly in nuclear power plants. Since the seized reactor coolant pump (RCP) accident of an advanced power reactor 1400 (APR1400) can be regarded as a representative core asymmetric event with respect to core inlet flow, the departure from nucleate boiling ratio (DNBR), which is a regulatory acceptance criterion in nuclear safety, should be evaluated with consideration of the uncertainty range of the core inlet flow reflecting the actual geometry. This study investigates the DNBR quantitatively in the entire fuel assemblies in the core using several codes for system behavior, computational flow dynamics, sub-channel analysis, and uncertainty evaluation. Based on the results from a system thermal-hydraulic analysis of a seized RCP accident of APR1400, this study presents the uncertainty range calculated by computational fluid dynamics on the asymmetry of the core inlet flow. Damaged fuel rods are quantitatively identified through a sub-channel analysis, which presents statistic relevance to obtain the DNBR at 95% reliability and 95% accuracy level. Additionally, an optimized evaluation methodology of a non-loss of coolant accident (non-LOCA) is realized by several nuclear codes.

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.

CHF prediction in rod bundles using round tube data

The present work concerns the use of 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, for the prediction of critical heat flux in rod bundles geometries. Comparisons between measured and calculated critical heat fluxes indicate that this table can be applied to rod bundles provided that a suitable correction factor is employed. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis.

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.

Sub-channel thermal-hydraulic analysis for VVER- 1000 generation III PWR

The sub-channel thermal-hydraulic analysis of a nuclear reactor is essential for assessing of its safety aspects. In this paper, the VVER-1000 has been selected as an example of the third generation reactors since it meets most of the international safety standards and because it has been taken as a base for designing the VVER-1200 which is belonging to the III+ generation. A steady state mathematical model has been proposed and solved to validate and assure that the hottest channel temperature limits are satisfied. The various temperature distributions, the critical heat flux and the departure from nucleate boiling ratio (DNBR) for the hottest channel were evaluated. Also, a transient state model has also been presented and solved using the finite difference method with the aid of MATLAB algorithm. An exponential loss of flow rate of the reactor core coolant was triggered from the steady state conditions. We assumed that the neutron flux and the generated power were unchanged during the postulated event. The average core coolant flow time constant was treated as a single parameter expressing the rapidity of the event. A value of 250 seconds time constant was assumed for slow transient, whereas 10 seconds was assumed for fast one. The reactor core was assumed to be protected through the reactor control system and mitigated according to the regular emergency operating procedures. The time dependent temperature distributions were calculated for the cladding of the hottest coolant channel. For each value of the temperature, the response time required for reaching unsafe conditions was evaluated, discussed and presented.

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.

Innovative designed fuel pin of conceptual semi-heavy water research reactor with improved reactor safety

Improving thermal hydraulic parameters such as axial power distribution and decrease maximum fuel temperatures lead to higher reactor safety. In this study, with the variation of fuel pellet central hole (1UTVS) volume along the fuel pin axis, have achieved improvements of thermal hydraulic parameters of the semi-heavy water research reactor. The new design of fuel pin has contain fuel pellets with different radii central hole along the fuel pin axis. Calculations with the new design of fuel pin are carried out by coupling of neutronic (WIMS-D5+CITATION) and thermal hydraulic (COBRA-EN) codes for a conceptual high neutron flux semi-heavy water research reactor. As a result, the axial Power Peaking Factor (PPF) decreased from 1.2827 to 1.2118 and minimum Departure from Nucleate Boiling Ratio (DNBR) has increased from 5.31 to 5.583. Maximum fuel temperature is decreased from 1595 to 1499.6 °C and core average thermal flux is increased from 1.1607 × 1014 to 1.2447 × 1014 n/cm2.s at the Beginning of Cycle (BOC). These improvement which are the consequence of applying the new innovative design of the fuel pin lead to higher safety. This new idea of designed fuel pin with different radii central hole fuel pellets along the fuel pin axis can be used for nuclear reactors in order to improve their safety.