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.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

Loss of Representativeness of the Fixed Incore Instrumentation of the EPR™ Reactor

2014 ◽  
Author(s):  
Brice Jardiné ◽  
Olivier Bougeant ◽  
Maxime Pfeiffer
Keyword(s):  
Power Distribution ◽  
Power Level ◽  
Nucleate Boiling ◽  
Inlet Temperature ◽  
Fuel Loading ◽  
Calculation Algorithm ◽  
The Core ◽  
On Line ◽  
The Difference ◽  
Online Reconstruction

The EPR™ reactor features a fixed incore instrumentation, composed of 72 Self Powered Neutron Detectors (SPND), that provides the online reconstruction of the core maximum Linear Power Density (LPD) and minimum Departure from Nucleate Boiling Ratio (DNBR). The Instrumentation and Control (I&C) systems of the EPR™ reactor use this online reconstruction in surveillance and protection functions. The onsite thresholds of those I&C functions have to take into account all the uncertainties affecting the online reconstruction of core power distribution measured by SPNDs. One of these uncertainties is the so-called Loss Of Representativeness (LOR). This uncertainty is defined as the difference between the LPD (respectively DNBR) physical value and the LPD (respectively DNBR) computed value using SPND signals. The LOR parameter is mostly linked to the difference between the core power distribution at the time where SPNDs are calibrated and the core power distribution at the time where their signals are used. For the DNBR, LOR also takes into account the use of a simplified on-line DNBR calculation algorithm. A statistical approach is used in order to define this uncertainty. The analysis is based on the evaluation of different sets of core power distributions generated thanks to random drawings of the plant state parameters (including power level, core inlet temperature, pressure, control rod insertion and xenon distribution). The sets of core configurations representative of normal plant operation are used to define the surveillance thresholds. The sets representative of accidental transients (for which the LPD and DNBR protections are claimed) are used to define the protection thresholds. The analysis of LOR values provides an envelop probability law covering a minimum of 95% of LOR values. In order to derive the on-site threshold for LPD and DNBR, a Monte Carlo method is used to propagate the LOR probability law and the other uncertainties. Sensitivity calculations have been performed in order to cover a large spectrum of fuel loading patterns and to take into account SPND failures. In conclusion, this approach allows defining an optimized and robust set of thresholds for the on-line surveillance and protection system of EPR™ reactor.

Numerical Simulation of Core Temperature and Melting Process of IVR Core After a Severe Water Loss Accident

2017 ◽  
Author(s):  
Yiqun Liu ◽  
Xiaoying Zhang ◽  
Jingya Li ◽  
Biao Wang ◽  
Dekui Zhan ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Water Level ◽  
Water Loss ◽  
Reactor Core ◽  
Melting Process ◽  
Nucleate Boiling ◽  
Maximum Temperature ◽  
Conduction Model ◽  
Fuel Rods ◽  
The Core

After the occurrence of severe water loss accident in a PWR, the water level in the reactor core would decrease gradually, leading to heat up and melted down of the core, threatening safety of the nuclear power plant and the surrounding environment. In this paper, the 1/4 core of AP1000 PWR was adopted for study, a numerical method has been established to calculate the transient change of temperature and melting process of the core and envelope structure (boarding, basket and RPV) after the severe water loss accident. A two-dimensional conduction model with cylindrical coordinate has been used to simulate heat transfer along the radius and height direction of fuel rods and control rods in fuel assemblies. Heat transfer condition on rod surface considers nucleate boiling for rod surface below the water level, while radiative heat transfer among neighboring rods and natural convection with water vapor was considered for rod surface above the water level. Heat transfer along thickness of envelope structures were modeled with the one-dimensional conduction model. The results show that the maximum temperature of the whole reactor core does not exceed 3000K and AP1000 will not meet the melting of fuel rods with the help of RPV external water chamber cooling. The temperature values of the fuel rods and the control rod show the characteristic distribution of the two regions. At 4904s, the maximum temperature of the rod rises to 2900K, and then stabilize. The temperature of the shell is up to 2000K, the maximum temperature of the basket is to 1260K, the variation of RPV wall temperature is not obvious.

Analysis of Thermal-Hydraulic Fluctuations in Trillo NPP With CTF/PARCS v. 2.7 Coupled Code

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
A. Abarca ◽  
R. Miró ◽  
G. Verdú ◽  
J. A. Bermejo
Keyword(s):  
Reactor Core ◽  
Low Frequency ◽  
Protection System ◽  
Inlet Temperature ◽  
Frequency Noise ◽  
Pressurized Water ◽  
The Core ◽  
Coupled Codes ◽  
Water Reactors ◽  
New Research

The low-frequency noises are fluctuations in the neutron flux density, in the low-frequency range up to 4 Hz, which generate noise in the neutron instrumentation and could affect the limitation and protection system of the reactor core. Some European pressurized water reactors (PWRs) experienced the effect of low-frequency noise, opening a new research line for the verification of the neutron-kinetics/thermal-hydraulic coupled codes. A CTF/PARCS v. 2.7 simulation study to verify whether periodical fluctuations in the core inlet temperature could activate the core protection system has been done, obtaining the frequency spectrum of the power oscillation amplitudes.

scholarly journals SPX Benchmark Part I: Results of Static Neutronics

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk ◽  
Liang Zhang ◽  
Evgeny Nikitin ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Thermal Conditions ◽  
Reaction Rates ◽  
Second Phase ◽  
Reference Solution ◽  
Transient Behaviour ◽  
The Core ◽  
Reactivity Coefficients

Abstract In the paper, the specification of a new neutronics benchmark for a large Sodium cooled Fast Reactor core and results of modelling by different participants are presented. The neutronics benchmark describes the core of the French sodium cooled reactor Superphénix at its startup configuration, which in particular was used for experimental measurement of reactivity characteristics. The benchmark consists of the detailed heterogeneous core specification for neutronic analysis and results of the reference solution. Different core geometries and thermal conditions from cold “as fabricated” up to full power were considered. The reference Monte Carlo solution of Serpent 2 includes data on multiplication factor, power distribution, axial and radial reaction rates distribution, reactivity coefficients and safety characteristics, control rods worth, kinetic data. The results of modelling with seven other solutions using deterministic and Monte Carlo methods are also presented and compared to the reference solution. The comparisons results demonstrate appropriate agreement of evaluated characteristics. The neutronics results will be used in the second phase of the benchmark for evaluation of transient behaviour of the core.

Mixing Conditions in the Lower Plenum and Core Inlet of a BWR

2008 ◽  
Author(s):  
Hernan Tinoco ◽  
Stefan Ahlinder
Keyword(s):  
Boundary Conditions ◽  
Three Dimensional ◽  
Mean Value ◽  
Large Temperature ◽  
Inlet Temperature ◽  
Inlet Section ◽  
The Core ◽  
Cfd Models ◽  
Stable Temperature ◽  
The One

A thermal mixing analysis of the Downcomer, Main Recirculation Pumps (MRPs) and Lower plenum of Forsmark’s Unit 3 has been carried out with three separate CFD models. Several difficulties with the boundary conditions have been encountered, particularly with the MRP model. The results obtained predict stable temperature differences of around 8 K at the core inlet. Such large temperature differences have never been observed at Forsmark NPP. Temperature measurements at four positions above the Reactor Pressure Vessel (RPV) bottom give the mean value used as the core inlet temperature for core analyses with codes such as POLCA. The temperature transmitters used are rather slow and inaccurate. Still, they should be able to detect large stable temperature differences such as those predicted by the aforementioned computations. Indirect indication of the incongruity of these predictions is the possibility of fuel damage caused by such large temperature differences. Fuel damage other than the one caused by debris fretting (thread-like particles) through mechanical influence has not been reported at Forsmark NPP since the implementation of liner cladding in fuel design. Also, the aforementioned difficulties with the connection of the models throw some doubt upon the accuracy of these predictions. A completely connected model of the same RPV volume covered by the separate models predicts temperature differences at core inlet that are almost a fourth of those mentioned above, i. e. approximately 2.5 K. Most of the mixing occur downstream of the MRP diffusers, at the Lower Plenum “inlet”. The reason for this prediction divergence is an impossibility of a correct transfer of complete three-dimensional flow field properties by means of boundary conditions defined at a two-dimensional inlet section.

Axial Offset Anomaly Modeling Studies on EDF Cores

2008 ◽  
Author(s):  
Etienne De´cossin ◽  
O̸ystein Bremnes
Keyword(s):  
Power Distribution ◽  
Single Phase ◽  
Nucleate Boiling ◽  
The United States ◽  
Fuel Surface ◽  
Root Cause ◽  
The Core ◽  
Visible Effect ◽  
Low Amplitude ◽  
Additional Constraints

The Axial Offset Anomaly phenomenon, commonly called AOA, is one of the possible consequences of the undesirable presence of deposits on the nuclear fuel. AOA appears in PWR cores as abnormal distribution of the power, as compared to the design reference values. If the amplitude of the phenomenon becomes significant, it may lead to additional constraints in operating the reactor. Several factors contribute to the root cause of AOA. The state-of-the art knowledge relies on the fact that deviation should appear when the conditions of temperature, boron and corrosion product concentrations are appropriate to form crud on the fuel surface, that is thick enough to allow precipitation of lithium borates. Then, the neutron capture is locally enhanced by the additional presence of boron, leading to local flux depression and redistribution of the power all over the core. The experience feedback on AOA is rather consequent, mainly in the United States. On the EDF nuclear fleet, excepted two notable cases in the mid 90’s, the phenomenon remained limited to a few, low amplitude observations on the 1300 MW type cores. To perform AOA studies, EDF has developed its own numerical tool based on: • the neutron kinetics COCCINELLE code for power distribution computations, • the thermal-hydraulics THYC code, • the dedicated BOA code to evaluate, at the core scale, crud deposition and boron loading. The EDF software COCCINELLE and THYC are commonly used for core design and safety analysis. For AOA studies, they provide a 3D, best-estimate representation of the clean core. Then the thermal-hydraulic data are used by the BOA code as boundary conditions to determine, in both single-phase flow and sub-cooled nucleate boiling, how and where deposits form on the fuel surface. In the standard approach, the total mass of trapped boron is compared to a threshold defining the AOA risk limit. In a prospective approach, an advanced COCCINELLE, THYC and BOA coupling is proposed to account for the power distribution changes all along the cycle, instead of using the clean core data. Numerical simulations of an AOA cycle of a 900 MW core show that this feedback effect has a visible effect on the final axial-offset prediction.

scholarly journals SPARK-NC: A Lead-Bismuth-Cooled Small Modular Fast Reactor with Natural Circulation and Load Following Capabilities

Energies ◽  
2020 ◽  
Vol 13 (20) ◽  
pp. 5410
Author(s):  
Muhammad Hashim ◽  
Liangzhi Cao ◽  
Shengcheng Zhou ◽  
Rubing Ma ◽  
Yiqiong Shao ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Natural Circulation ◽  
Conservation Equation ◽  
Reactor Power ◽  
Inlet Temperature ◽  
Primary Control ◽  
The Core ◽  
Load Following ◽  
Energy Conservation Equation

In this study, a conceptual design was developed for a lead-bismuth-cooled small modular fast reactor SPARK-NC with natural circulation and load following capabilities. The nominal rated power was set to 10 MWe, and the power can be manipulated from 5 MWe to 10 MWe during the whole core lifetime. The core of the SPARK-NC can be operated for eight effective full power years (EFPYs) without refueling. The core neutronics and thermal-hydraulics design calculations were performed using the SARAX code and the natural circulation capability of the SPARK-NC was investigated by employing the energy conservation equation, pressure drop equation and quasi-static reactivity balance equation. In order to flatten the radial power distribution, three radial zones were constructed by employing different fuel enrichments and fuel pin diameters. To provide an adequate shutdown margin, two independent systems, i.e., a control system and a scram system, were introduced in the core. The control assemblies were further classified into two types: primary control assemblies used for reactivity control and power flattening and secondary control assemblies (with relatively smaller reactivity worth) used for power regulation. The load following capability of SPARK-NC was assessed using the quasi-static reactivity balance method. By comparing three possible approaches for adjusting the reactor power output, it was shown that the method of adjusting the coolant inlet temperature was viable, practically easy to implement and favored for the load following operation.

Mixing Conditions in the Lower Plenum and Core Inlet of a Boiling Water Reactor

2009 ◽  
Vol 131 (6) ◽  
Author(s):  
Hernan Tinoco ◽  
Stefan Ahlinder
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Mean Value ◽  
Large Temperature ◽  
Inlet Temperature ◽  
Inlet Section ◽  
The Core ◽  
Fuel Design ◽  
Dynamics Models

A thermal mixing analysis of the downcomer, main recirculation pumps (MRPs) and lower plenum of Forsmark’s Unit 3 has been carried out with three separate computational fluid dynamics models. Several difficulties with the boundary conditions have been encountered, particularly with the MRP model. The results obtained predict stable temperature differences of around 8 K at the core inlet. Such large temperature differences have never been observed at Forsmark nuclear power plant (NPP). Temperature measurements at four positions above the reactor pressure vessel (RPV) bottom give the mean value used as the core inlet temperature for core analyses. Even if the temperature transmitters used are rather slow and inaccurate, they should be able to detect such large temperature differences that may lead to fuel damage. The only damage reported at Forsmark NPP since the implementation of liner cladding in fuel design is that caused by mechanically induced debris fretting (threadlike particles). Also, the difficulties with the connection of the models throw some doubt on the accuracy of these predictions. A completely connected model of the same RPV volume covered by the separate models predicts temperature differences at core inlet that are almost one-fourth of those mentioned above, i.e., approximately 2.5 K. Most of the mixing occurs downstream of the MRP diffusers, at the lower plenum “inlet.” This prediction divergence seems to arise from an impossibility of a correct transfer of complete three-dimensional flow field properties by means of boundary conditions defined at a two-dimensional inlet section.

scholarly journals Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

2014 ◽  
Vol 2014 ◽  
pp. 1-8
Author(s):  
Po Hu ◽  
Paul P. H. Wilson
Keyword(s):  
Steady State ◽  
Supercritical Water ◽  
Power Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Core Model ◽  
Core Flow ◽  
The Core ◽  
Extended Code ◽  
Reference Design

This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

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