scholarly journals Verification of SARAX code system in the reactor core transient calculation based on the simplified EBR-II benchmark

2021 ◽  
Author(s):  
Xiaoqian Jia ◽  
Youqi Zheng ◽  
Xianna Du ◽  
Yongping Wang ◽  
Jianda Chen
Keyword(s):  
Reactor Core ◽  
Code System

VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

2016 ◽  
Author(s):  
Vefa N. Kucukboyaci ◽  
Yixing Sung ◽  
Yiban Xu ◽  
Liping Cao ◽  
Robert K. Salko
Keyword(s):  
Reactor Core ◽  
Steam Line ◽  
High Flow ◽  
Low Flow ◽  
Code System ◽  
Fuel Temperature ◽  
Transport Code ◽  
Main Steam Line ◽  
Main Steam ◽  
Flow Case

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics (T/H), and fuel temperature components with an isotopic depletion capability. The neutronics capability is based on the Michigan Parallel Characteristics Transport Code (MPACT), a three-dimensional whole-core transport code. The T/H and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to the departure from nucleate boiling (DNB) ratio at the most limiting point of a postulated pressurized water reactor main steam line break event initiated at the hot zero power, either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power, where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady-state reactor core response under the main steam line break accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on the HTR-10

2008 ◽  
Author(s):  
Yujie Dong ◽  
Fubing Chen ◽  
Zuoyi Zhang ◽  
Shouyin Hu ◽  
Lei Shi ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Power Level ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Code System ◽  
Safety Features ◽  
Forced Cooling ◽  
High Temperature Gas

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

scholarly journals Verification of a two-step code system MCS/RAST-F to fast reactor core analysis

2021 ◽  
Author(s):  
Tuan Quoc Tran ◽  
Alexey Cherezov ◽  
Xianan Du ◽  
Deokjung Lee
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Core Analysis ◽  
Code System

scholarly journals CALCULATION OF 2-DIMENSIONAL PWR MOX/UO2 CORE BENCHMARK OECD NEA 6048 WITH SRAC CODE

2020 ◽  
Vol 22 (3) ◽  
pp. 89
Author(s):  
Wahid Luthfi ◽  
Surian Pinem
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Design Parameters ◽  
Maximum Deviation ◽  
Maximum Difference ◽  
Code System ◽  
Spent Fuel ◽  
Neutronic Parameters ◽  
Safety Characteristic

The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation of cross-sections and calculation of reactor core design parameters such as criticality, reactivity of control rods and radial power distribution. In this study, PWR MOX/UO2 Core Transient Benchmark is used to verify the code that models a MOX/UO2 fueled core. SRAC-CITATION result is different from DeCART by 0.339% from. SRAC-CITATION result of single rod worth in All Rods Out (ARO) conditions are quite good with a maximum difference of 6.34% compared to BARS code and 4.74% compared to PARCS code. In All Rods In (ARI) condition, SRAC-CITATION results compared to the PARCS code is quite good where the maximum difference is 9.72%, but compared to BARS code, it spikes up to 33.24% at maximum difference. In the other case, overall radial power density results are quite good compared to the reference. Its maximum deviation from DeCART code is 5.325% in ARO condition and 6.234% in ARI condition. Based on the results of these calculations, SRAC code system can be used to generate cross-section and to calculate some neutronic parameters. Hence, it can be used to evaluate the neutronic parameters of the MOX/UO2 PWR core design.Keywords: MOX/UO2 fuel, Criticality, Power peaking factor, SRAC2006

scholarly journals VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK

2021 ◽  
Vol 27 (1) ◽  
pp. 47
Author(s):  
Wahid Luthfi ◽  
Surian Pinem
Keyword(s):  
Boron Concentration ◽  
Reactor Core ◽  
Axial Direction ◽  
Code System ◽  
Pressurized Water ◽  
Calculation Results ◽  
Neutronic Parameters ◽  
Weighted Error ◽  
Parameters Calculation

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction, and Power Peaking Factor (PPF) and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel.Keywords: Neutronic parameter, critical boron concentration, power peaking factor, SRAC code system.

scholarly journals OSCAR-4 CODE SYSTEM COMPARISON AND ANALYSIS WITH A FIRST-ORDER SEMI-EMPIRICAL METHOD FOR CORE-FOLLOW DEPLETION CALCULATION IN MNR

2020 ◽  
Vol 9 (1) ◽  
pp. 73-82
Author(s):  
Mohammed Alqahtani ◽  
Simon Day ◽  
Adriaan Buijs
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Empirical Method ◽  
Code System ◽  
First Order ◽  
Nuclear Reactor Fuel ◽  
Nuclear Reactor Core ◽  
Comparison And Analysis ◽  
Operating Reactor ◽  
Semi Empirical

Knowledge of the isotopic composition of a nuclear reactor core is important for accurate core-follow and reload analysis. In the McMaster Nuclear Reactor, fuel depletion estimates are based upon a semi-empirical calculation using flux-wire measurements. These estimates are used to plan and guide fuelling operations. To further support operations, an OSCAR-4 model is being developed. To evaluate the performance of the OSCAR-4 code for this application, 2 points of comparison, considering the period between 2007 and 2010, are presented: (i) the multiplication factor keff and (ii) U-235 fuel inventory. The latter is compared with a simple first-order semi-empirical calculation. The calculation of keff for the last operational 3 months yields 0.997 ± 0.002 (vs. 1.000 for an operating reactor), and differences in both core-average inventory and the maximum standard fuel assembly inventories estimates are found to be 5.7% and 7.5%, respectively.

scholarly journals Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

2004 ◽  
Vol 1 (3) ◽  
pp. 99-112
Author(s):  
Vladan Ljubenov ◽  
Miodrag Milosevic
Keyword(s):  
Neutron Flux ◽  
Cross Section ◽  
Reactor Core ◽  
Numerical Procedure ◽  
Experimental Methodology ◽  
Neutron Absorption ◽  
Code System ◽  
Neutron Converter ◽  
Neutron Absorption Cross Section

The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114.

Methods for Modeling Local Fuel Rod Crud and Assessing Failure Risk

2009 ◽  
Author(s):  
Michael A. Krammen ◽  
Guoqiang Wang ◽  
Steven F. Grill ◽  
Zeses E. Karoutas ◽  
Michael Y. Young
Keyword(s):  
Reactor Core ◽  
Thermal Conditions ◽  
Code System ◽  
Key Factors ◽  
Reactor Coolant ◽  
Fuel Rod ◽  
Hydraulic Conditions ◽  
Fuel Design ◽  
Computational Fluid Dynamics Cfd ◽  
Deposition Models

A code system with attendant methods has been developed for modeling local fuel rod crud. The methodology is in production use, when appropriate, in 14×14 and 16×16 CE-design fuel PWR reloads. This tool is used in making EPRI Level IV type crud and corrosion guideline assessments, where the EPRI crud and corrosion guidelines were developed in response to the INPO zero fuel failures by 2010 initiatives. The methodology is in process of being extended to other Westinghouse fuel designed reloads. The methodology involves translating very detailed relative thermal hydraulic variations determined by Computational Fluid Dynamics (CFD) computations over a fuel assembly grid span into specific local thermal hydraulic conditions over the entire axial length of every fuel rod in a reactor core over the life of the rod in reactor. The very local thermal hydraulic conditions are combined with reactor coolant crud concentrations as input to models for predicting very local fuel rod crud deposition. The reactor coolant crud concentrations are determined over each reactor cycle by reactor system wide crud mass balance calculations. The reactor coolant crud concentrations are used to calculate local crud thickness using mass transfer models which are a function of the local thermal conditions. The advanced crud deposition models also include models for calculating local crud dryout. Local crud deposition and crud dryout are strongly dependent on very local boiling or steaming which are predicted through the translation of the CFD results. The local crud thickness and degree of local crud dryout are key factors in determining the margin or risk for local fuel rod cladding crud induced fuel failure. Guideline limits were established by benchmarking the methodology to recent fuel rod crud induced fuel failures. This paper presents details on each facet of this methodology. Results for a particular application are provided to illustrate the methodology. The application is for a plant where a new fuel design was implemented with mixing grids and intermediate flow mixers. The results of the analysis demonstrated acceptable levels of local crud thickness and percent local crud dryout.

Point Defect Absorption and Dislocation Loop Formation at Grain Boundaries in Hvem-Irradiated Zr and Its Alloys:1. Influence of Solute Addition

1985 ◽  
Vol 43 ◽  
pp. 350-351
Author(s):  
R.A. Herring ◽  
M. Griffiths ◽  
M.H Loretto ◽  
R.E. Smallman
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Point Defect ◽  
Reactor Core ◽  
Solute Concentration ◽  
Nuclear Industry ◽  
Dislocation Loops ◽  
Loop Formation ◽  
Component Material ◽  
Impurity Interaction

Because Zr is used in the nuclear industry to sheath fuel and as structural component material within the reactor core, it is important to understand Zr's point defect properties. In the present work point defect-impurity interaction has been assessed by measuring the influence of grain boundaries on the width of the zone denuded of dislocation loops in a series of irradiated Zr alloys. Electropolished Zr and its alloys have been irradiated using an AEI EM7 HVEM at 1 MeV, ∼675 K and ∼10-6 torr vacuum pressure. During some HVEM irradiations it has been seen that there is a difference in the loop nucleation and growth behaviour adjacent to the grain boundary as compared with the mid-grain region. The width of the region influenced by the presence of the grain boundary should be a function of the irradiation temperature, dose rate, solute concentration and crystallographic orientation.

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