Margin Assessment of AP1000 Loss of Flow Transient

2006 ◽  
Author(s):  
Edward L. Carlin ◽  
Peter A. Hilton ◽  
Yixing Sung
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Improved Method ◽  
Margin Assessment ◽  
Reactor Coolant ◽  
Power Profile ◽  
Canned Motor ◽  
Reactivity Insertion ◽  
Coolant System

The Reactor Coolant System (RCS) of the AP1000 plant consists of two circulating loops. Each loop contains two canned motor Reactor Coolant (RC) pumps that have a rotating inertia to provide RCS flow coastdown if power to the pumps is lost. Westinghouse analysis of the complete loss of flow (CLOF) accident in support of the AP1000 design certification was based on the USNRC-approved traditional methodology applied to operating plants. The RCS response during the transient was predicted using the LOFTRAN code based on a reactivity insertion curve highly skewed to the bottom of the reactor core, but the calculation of Departure from Nucleate Boiling Ratio (DNBR) was performed assuming a top-skewed axial power profile. A more realistic margin assessment can be made by using an improved method similar to Westinghouse RAVE methodology recently approved by the USNRC. The improved method uses the three-dimensional kinetic nodal code SPNOVA coupled with the reactor core thermal-hydraulic code VIPRE-W for predicting the reactor core response during the CLOF transient. The improved method significantly improves margin predictions by generating core power distributions consistent with the trip reactivity changes for the DNBR calculation. The margin assessment showed that the improved method resulted in a 19% DNBR increase as compared to the traditional method for the AP1000 CLOF transient.

Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe

2014 ◽  
Author(s):  
Bo Shi ◽  
Zhao-Fei Tian
Keyword(s):  
Real Time ◽  
Data Exchange ◽  
Reactor Core ◽  
Simulation Software ◽  
Modeling Method ◽  
Time Step ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Coolant System

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

scholarly journals Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 531-536 ◽  
Author(s):  
Igor P. Królikowski ◽  
Jerzy Cetnar
Keyword(s):  
Cross Sections ◽  
Nuclear Reactions ◽  
Nuclear Reactors ◽  
Input Parameter ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled Analysis ◽  
Thermal Hydraulics ◽  
Power Profile ◽  
The Impact

Abstract Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

MTC 3D: A Fully Coupled Methodology for MSLB Safety Analysis

2014 ◽  
Author(s):  
Christian Royère ◽  
Michel Gonnet ◽  
Brahim Manchid
Keyword(s):  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Steam Line ◽  
Second Step ◽  
Inlet Temperature ◽  
Original Design ◽  
The Core ◽  
Strong Asymmetry

The main steam line break (MSLB) is an overcooling accident that may lead to an over criticality, and so to a power increase, after the reactor trip. The most penalizing single failure is a RCCA bank stuck out of the core when the reactor trip occurs. This configuration leads to a strong asymmetry of the radial power shape combined with a strong asymmetry of the core inlet temperature that results in a strongly distorted 3D power distribution. In the original design, the MSLB accident was studied with a simplified and conservative 0D method. The point kinetics approach requires the use of extremely conservative assumptions in order to account for the asymmetry in the core region that takes place during the transient. The use of the coupling between a three-dimensional neutronic code (SMART), a 3D core thermal-hydraulic code (FLICA cf. ref [4]) and a reactor coolant system code (MANTA cf. ref [3]) allows representing the 3D heterogeneity of the power shape and also of the resulting cross flows. In addition, this coupling allows determining moderator and Doppler feedback effects in a much more realistic way thus limiting accident consequences estimated. A methodology, called MTC3D (for Méthode Totalement Couplée 3D in French), has been developed using the coupling between the three codes to perform the MSLB analysis. The physical dominant parameters of the transient are identified through a comprehensive sensitivity analysis. Then, a deterministic approach is used in the entire transient simulation considering dominant parameters in a penalizing way. In a first step, neutronic data are determined with SMART calculations. In a second step, MANTA/SMART/FLICA transients are performed with penalized neutronic and thermal-hydraulic data. In a third step, as the steam line break transient is a relatively slow transient, the core power distribution is evaluated with a steady state SMART/FLICA calculation without penalization. In a last step, safety criteria, such as minimum DNBR (Departure from Nucleate Boiling Ratio) are calculated with FLICA calculations based on core power distribution calculated at the third step and boundary conditions calculated at the second step. The use of 3D neutronic and detailed thermal-hydraulic codes to model the reactor core allows considering a more physical representation of the core configuration for transient analysis. The coupling between 3D neutronic and core thermal-hydraulic codes allows exhibiting intrinsic margins without over penalizations related to a simplified 0D method.

Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

Kerntechnik ◽  
2022 ◽  
Vol 0 (0) ◽  
Author(s):  
Alexandre de Souza Soares ◽  
Antonio C. M. Alvim
Keyword(s):  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Operating Personnel ◽  
Reactor Coolant ◽  
Operational Variables ◽  
Emergency Power ◽  
Power Mode ◽  
Reactor Shutdown ◽  
Coolant System

Abstract The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

Thermodynamic Data of Iodine Reactions Calculated by Quantum Chemistry. Training Set of Molecules

2008 ◽  
Vol 73 (10) ◽  
pp. 1340-1356 ◽  
Author(s):  
Katarína Mečiarová ◽  
Laurent Cantrel ◽  
Ivan Černušák
Keyword(s):  
Quantum Chemistry ◽  
Ab Initio ◽  
Thermodynamic Data ◽  
Theoretical Calculations ◽  
Reactor Core ◽  
Fission Products ◽  
Vapour Phase ◽  
Reactor Accident ◽  
Training Set ◽  
Coolant System

This paper focuses on the reactivity of iodine which is the most critical radioactive contaminant with potential short-term radiological consequences to the environment. The radiological risk assessments of 131I volatile fission products rely on studies of the vapour-phase chemical reactions proceeding in the reactor coolant system (RCS), whose function is transferring the energy from the reactor core to a secondary pressurised water line via the steam generator. Iodine is a fission product of major importance in any reactor accident because numerous volatile iodine species exist under reactor containment conditions. In this work, the comparison of the thermodynamic data obtained from the experimental measurements and theoretical calculations (approaching "chemical accuracy") is presented. Ab initio quantum chemistry methods, combined with a standard statistical-thermodynamical treatment and followed by inclusion of small energetic corrections (approximating full configuration interaction and spin-orbit effects) are used to calculate the spectroscopic and thermodynamic properties of molecules containing atoms H, O and I. The set of molecules and reactions serves as a benchmark for future studies. The results for this training set are compared with reference values coming from an established thermodynamic database. The computed results are promising enough to go on performing ab initio calculations in order to predict thermo-kinetic parameters of other reactions involving iodine-containing species.

Numerical and Experimental Research on the Fluid-Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Rui Xu ◽  
Yun Long ◽  
Yaoyu Hu ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Cfd Method ◽  
Canned Motor

Reactor coolant pump (RCP) is one of the most important equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor RCP, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a long clearance flow. The fluid-induced forces of the clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally analyzed in this work. A transient computational fluid dynamics (CFD) method has been used to investigate the fluid-induced force of the clearance. A vertical experiment rig has also been established for the purpose of measuring the fluid-induced forces. Fluid-induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the CFD method and the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor RCP does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid-induced forces of the clearance flow.

Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

2018 ◽  
Author(s):  
Yuchuan Guo ◽  
Guanbo Wang ◽  
Dazhi Qian ◽  
Heng Yu ◽  
Bo Hu
Keyword(s):  
Fuel Assembly ◽  
System Analysis ◽  
Research Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Regulation System ◽  
Lower Power ◽  
Emergency Shutdown ◽  
Flow Blockage

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

Criticality Calculations and Basic Sensitivity/Uncertainty Investigation of LR-0 Benchmark Core

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Tomáš Czakoj ◽  
Evžen Losa
Keyword(s):  
Lithium Fluoride ◽  
Three Dimensional ◽  
Reactor Core ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Multiplication Factor ◽  
Reference Case ◽  
Energy Version ◽  
Data Libraries ◽  
Nuclear Data Libraries

Three-dimensional Monte Carlo code KENO-VI of SCALE-6.2.2 code system was applied for criticality calculation of the LR-0 reactor core. A central module placed in the center of the core was filled by graphite, lithium fluoride-beryllium fluoride (FLIBE), and lithium fluoride-sodium fluoride (FLINA) compounds. The multiplication factor was obtained for all cases using both ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. Obtained results were compared with benchmark calculations in the MCNP6 using ENDF/B-VII.0 library. The results of KENO-VI calculations are found to be in good agreement with results obtained by the MCNP6. The discrepancies are typically within tens of pcm excluding the case with the FLINA filling. Sensitivities and uncertainties of the reference case with no filling were determined by a continuos-energy version of the TSUNAMI sequence of SCALE-6.2.2. The obtained uncertainty in multiplication factor due to the uncertainties in nuclear data is about 650 pcm with ENDF/B-VII.1.

Sign in / Sign up

Export Citation Format

Share Document