scholarly journals Reactivity insertion accident analysis during uranium foil target irradiation in the RSG-GAS reactor core

2020 ◽  
Vol 35 (3) ◽  
pp. 201-207
Author(s):  
Surian Pinem ◽  
Tagor Sembiring ◽  
Tukiran Surbakti
Keyword(s):  
Steady State ◽  
Reactor Core ◽  
Maximum Temperature ◽  
Diffusion Method ◽  
Accident Analysis ◽  
Calculation Results ◽  
Foil Target ◽  
Enriched Uranium ◽  
Initial Power ◽  
Reactivity Insertion

Analysis of the steady-state and reactivity insertion accident is very important for the safety of reactor operations. In this study, steady-state and reactivity insertion accident analysis when the low enriched uranium foil target is irradiated in the reactor core has been carried out. The analysis is carried out by the best estimate method by using a coupled neutronic, kinetic, and thermal-hydraulic code, MTR-DYN. The MTR-DYN code is based on the 3-D multigroup neutron diffusion method. The cell calculations for the target are carried out by the WIMSD/5 and MTR-DYN code. After reactivity insertion, the coolant, fuel, and clad temperature are observed. The calculation results for the initial power of 1 W showed that the maximum temperature of the coolant, clad, and fuel are 49.76?C, 65.01?C, and 65.26?C, respectively. Meanwhile, when the reactivity insertion at the initial power of 1 MW, the maximum temperature of the coolant, clad, and fuel are 72.23?C, 140.79?C, and 141.97?C, respectively. Based on those calculation results during irradiation low enriched uranium foil target, the temperature in the steady-state and reactivity insertion accident does not exceed the allowable safety limit.

Thermal-Hydraulic Analysis of TOPAZ-II With Modified RELAP5

2018 ◽  
Author(s):  
Simiao Tang ◽  
Chenglong Wang ◽  
G. H. Su ◽  
Suizheng Qiu ◽  
Wenxi Tian
Keyword(s):  
Steady State ◽  
Nuclear Power ◽  
Power Reactor ◽  
Liquid Metals ◽  
High Reliability ◽  
Maximum Temperature ◽  
Nuclear Power Reactor ◽  
Reactor Kinetics ◽  
Space Nuclear Power ◽  
Reactivity Insertion

With the advantages of high reliability, high power density and long life, small nuclear power reactor has become one of the most excellent space power options in the space missions. TOPAZ-II is the most mature space nuclear power reactor based on thermionic conversion. In this paper, the thermo-physical and transport properties of NaK-78 and heat transfer correlations for liquid metals are implemented into the RELAP5 code. The modified RELAP5 has already been accessed to analyze the thermal-hydraulic characteristics of the space reactor cooled by NaK-78. A RELAP5 model including the core, TFEs, radiator, coolant loop and volume accumulator is developed. Temperature reactivity feedback, TFE emitter, TFE collector, moderator and the reactivity insertion effects of control drums and safety drums are modeled in the point reactor kinetics equations with six-group delayed neutrons. To V&V the integrated TOPAZ-II system model, the steady state is simulated and analyzed. The steady state calculated results are in good agreement with the designed values. On the basis of V&V, a hypothetical reactivity insertion accident is simulated and analyzed. During the accident, the automatic control system is assumed to be malfunctioned, 0.01$ positive reactivity is introduced for 500s and then control drums start to rotate inward. The maximum temperatures of fuel and emitter are below the melting temperature, respectively. The maximum temperature of coolant is 940K with 160K margin from boiling. With the rotating of control drums, the reactor reaches critical again.

scholarly journals Evaluation Of Radionuclides Release Estimation Of Power Reactor Using Scdap / Relap

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Jupiter Sitorus Pane
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Fuel Temperature ◽  
Primary Coolant ◽  
Calculation Results ◽  
Computer Codes ◽  
Radionuclide Release ◽  
Cladding Material

<p>Incident of radiation release to the environment is important event in reactor safety analysis. Numerous studies have been conducted using various computer codes, including SCDAP/RELAP, to calculate radionuclide releases into the reactor coolant during severe accident. This paper contains description of calculation results of radionuclide release from reactor core to primary coolant system in a1000 MW PWR reactor with the aim to study behavior of radionuclide releases during severe accident. The calculations using SCDAP/RELAP was done by assuming that there has been a station black out which ends up with some vapor released into the containment. As a result, the water level in core was reduced up to a level where the core is no longer covered by water. The uncovered core heats up to certain temperature where the oxidation of the cladding started to occur.  Afterwards the oxidation generated heat made fuel melting temperature reached and as consequences the release of radionuclide to the primary coolant.  The calculations show that in parallel with the increasing of fuel temperature, the radionuclide releases into the gap through diffusion started at time of 2000seconds after initial simulation but with a neglected concentration. Subsequently at the time of 29200seconds, the temperature reached more than 1000 K and the oxidation of the Zr-cladding material occurred which accelerated the fuel temperature increase and as well as radionuclide release. At34000seconds, maximum temperature of core reached 2800 K and radionuclide release into the primary cooling system started. At this time, accumulated dissolve fission product reached amount of 74.5 kg, while the non-condensable radionuclide reached 122 kg. However, these value need to be investigated further.</p>

Investigation on Coupling Behaviors of Thermal-Hydraulics/Neutronics Under Asymmetrical Inlet Conditions

2014 ◽  
Author(s):  
Yong Zheng ◽  
Min-jun Peng ◽  
Geng-lei Xia ◽  
Ren Li
Keyword(s):  
Steady State ◽  
Phase Ii ◽  
Reactor Core ◽  
Diffusion Method ◽  
Transient Condition ◽  
Computational Techniques ◽  
Thermal Hydraulics ◽  
The Core ◽  
State Calculation ◽  
Inlet Conditions

The reactor core is a complex system involving the reactor physics, thermal hydraulics and many other aspects. That means the distribution of the core power largely determines the profile of the thermal parameters, meanwhile the local thermal-hydraulics condition will in turn affect the neutronics calculation by moderator temperature effect and Doppler effects. Issues coupling the thermal-hydraulics with neutronics of nuclear plants still challenge the design, safety and the operation of LWR few years ago. Fortunately, the recent availability of powerful computer and computational techniques has enlarged the capabilities of making more realistic simulations of complex phenomena in NPPs. The current study deals with the development of an integrated thermal-hydraulics/neutronics model for Qinshan phase II NPP project reactor for the analysis of specific plant transients in which the neutronic response of the core is important, application of RELAP5-HD making use of the Helios code to derive the macroscopic cross-sections. Based on the coupled model, the steady state calculation and the transient simulation, involving the abnormal operation mode with asymmetrical coolant flux and temperature on the inlet of reactor, have been performed. The results show that the values obtained from coupled code RELAP5-HD calculation are in good agreement with the available experimental data, and the calculated accident parameters curves can predict all major trends of the transient. Steady state and transient condition calculation results are in accordance with the theoretical analysis from the aspect of coupled thermal-hydraulics/neutronics, this demonstrated a successful best estimate coupled RELAP5-HD model of Qinshan phase II NPP reactor has been developed, and the established model will provide a good foundation for the further analysis of the primary loop. It also can be concluded that the more accurate CFD method coupling three dimensional neutron kinetics code based on neutron diffusion method are necessary for steady-state calculation and analysis of transient/accident conditions when asymmetrical processes take place in the core. It is worth mentioning that RELAP5-HD code has already programmed the human-machine interface and the interface for coupling with other code, hence RELAP5-HD code has a broad application prospect in PWRs safety analysis.

scholarly journals ANALYSIS OF REACTIVITY INSERTION AS A FUNCTION OF THE RSG-GAS FUEL BURN-UP

2021 ◽  
Vol 23 (1) ◽  
pp. 1
Author(s):  
Tukiran Surbakti ◽  
Surian Pinem ◽  
Lily Suparlina
Keyword(s):  
Reactor Core ◽  
Static Condition ◽  
Power Coefficient ◽  
Maximum Temperature ◽  
Control Rod ◽  
Chain Reactions ◽  
The Core ◽  
Fuel Burn ◽  
Reactivity Insertion ◽  
Safety Limit

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.

scholarly journals REACTIVITY INSERTION ACCIDENT ANALYSIS OF KARTINI REACTOR

2011 ◽  
Vol 14 (2) ◽  
Author(s):  
Tegas Sutondo
Keyword(s):  
Steady State ◽  
Temperature Increase ◽  
Transient Analysis ◽  
Power Level ◽  
Transient Condition ◽  
Accident Analysis ◽  
Fuel Temperature ◽  
Severe Condition ◽  
Start Up ◽  
Reactivity Insertion

REACTIVITY INSERTION ACCIDENT ANALYSIS OF KARTINI REACTOR. A transient analysis of reactivity insertionaccident of Kartini reactor during start up from the minimum critical condition has been performed to estimate the effect onthe fuel temperature increase. Two cases of reactivity insertion limits had been assumed in this study i.e. the reactivityinsertions were limited by the actuation of overpower trip system (110 %) for the 1st case and by manual scram when thecontrol rod reached the 100 % UP position, assuming the overpower trip system was failed to function for the 2nd case.Adiabatic condition was assumed in this study, to get the most severe condition. The result shows that based on theassumed power level of trip setting for the 1st case, the total reactivity insertion was 0.82 $, corresponding to the reactorperiod of about 2 s and causing the maximum fuel temperature increase of around 11 oC or the maximum fuel temperature of45 oC. For the 2nd case the total reactivity insertion at the trip point was 1.367 $, resulting in the maximum fuel temperatureincrease of about 103 oC or maximum fuel temperatur of around 137 oC which is still far below the defined design limit of1115 oC for transient condition and 700 oC for steady state as well. This result concludes that by limiting the available coreexcess of reactivity at reasonably low, it could prevent the fuel from possible of undergoing an excessive temperatureincrease, during the postulated reactivity insertion accident.Keywords: Transient analysis, reactivity insertion, accident, reactor kartini, fuel temperature.

scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Hiroshi Madokoro ◽  
Alexei Miassoedov ◽  
Thomas Schulenberg
Keyword(s):  
Structural Analysis ◽  
Message Passing ◽  
Message Passing Interface ◽  
Severe Accident ◽  
Coupled Analysis ◽  
Accident Analysis ◽  
Analysis Tool ◽  
Lower Head ◽  
Calculation Results ◽  
Reactor Pressure

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

Preliminary Core Design and Optimization Analysis of Travelling Wave Reactor

2014 ◽  
Vol 1070-1072 ◽  
pp. 357-360
Author(s):  
Dao Xiang Shen ◽  
Yao Li Zhang ◽  
Qi Xun Guo
Keyword(s):  
Monte Carlo ◽  
Rational Design ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Optimization Analysis ◽  
Design And Optimization ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Ignition Zone

A travelling wave reactor (TWR) is an advanced nuclear reactor which is capable of running for decades given only depleted uranium fuel, it is considered one of the most promising solutions for nonproliferation. A preliminary core design was proposed in this paper. The calculation was performed by Monte Carlo method. The burning mechanism of the reactor core design was studied. Optimization on the ignition zone was performed to reduce the amount of enriched uranium initially deployed. The results showed that the preliminary core design was feasible. The optimization analysis showed that the amount of enriched uranium could be reduced under rational design.

Transient Thermoelastohydrodynamic Study of Tilting-Pad Journal Bearings Under Dynamic Loading

1998 ◽  
Vol 120 (2) ◽  
pp. 405-409 ◽  
Author(s):  
P. Monmousseau ◽  
M. Fillon ◽  
J. Freˆne
Keyword(s):  
Steady State ◽  
Dynamic Loading ◽  
Dynamic Load ◽  
Journal Bearings ◽  
Operating Conditions ◽  
Maximum Temperature ◽  
Tilting Pad ◽  
Transient Period ◽  
Final Steady State ◽  
Tilting Pad Journal Bearings

Nowadays, tilting-pad journal bearings are submitted to more and more severe operating conditions. The aim of this work is to study the thermal and mechanical behavior of the bearing during the transient period from an initial steady state to a final steady state (periodic). In order to study the behavior of this kind of bearing under dynamic loading (Fdyn) due to a blade loss, a nonlinear analysis, including local thermal effects, realistic boundary conditions, and bearing solid deformations (TEHD analysis) is realized. After a comparison between theoretical results obtained with four models (ISO, ADI, THD, and TEHD) and experimental data under steady-state operating conditions (static load Ws), the evolution of the main characteristics for three different cases of the dynamic load (Fdyn/Ws < 1, Fdyn/Ws = 1 and Fdyn//Ws > 1) is discussed. The influence of the transient period on the minimum film thickness, the maximum pressure, the maximum temperature, and the shaft orbit is presented. The final steady state is obtained a long time after the appearance of a dynamic load.

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