scholarly journals Analysis and comparison of computer programs to analyze the irradiation performance of Uranium Molybdenum monolithic fuel plates and Uranium dioxide cylindrical fuel rods in power reactors.

2021 ◽  
Vol 9 (1) ◽  
Author(s):  
André Luiz Candido da Silva ◽  
Antonio Teixeira e Silva
Keyword(s):  
Uranium Dioxide ◽  
Nuclear Power ◽  
Reactor Core ◽  
Mechanical Analysis ◽  
Operating Conditions ◽  
Nuclear Power Reactor ◽  
Fuel Rods ◽  
Light Water ◽  
Higher Power ◽  
Irradiation Performance

The aim of this work is to present a comparative analysis in terms of the irradiation performance of cylindrical uranium dioxide fuel rods and monolithic uranium molybdenum fuel plates in pressurized light water reactors.To analyze the irradiation performance of monolithic uranium molybdenum fuel plates when subjected to steady state operating conditions in light water pressurized reactors, the computer program PADPLAC-UMo was used, which performs thermal and mechanical analysis of the fuel taking into account the physical , chemicals and irradiation effects to which this fuel is subjected. For the analysis of the uranium dioxide fuel rods, the code FRAPCON was used, which is an analytical tool that verifies the irradiation performance of fuel rods of pressurized light water reactor, when the power variations and the boundary conditions are slow enough for the term permanent regime to be applied. The analysis for a small nuclear power reactor, despite the higher power density applied to the fuel plate in relation to the fuel rod, showed that the fuel plates have lower temperatures and lower fission gas releases throughout the analyzed power history, allowing the use of a more compact reactor core without exceeding the design limits imposed on nuclear fuel.

scholarly journals Temperature Dependence of Primary Species G(values) Formed from Radiolysis of Water by Interaction of Tritium β-Particles

2017 ◽  
Vol 19 (1) ◽  
pp. 17
Author(s):  
Sofia Loren Butarbutar ◽  
Sriyono Sriyono ◽  
Geni Rina Sunaryo
Keyword(s):  
Temperature Dependence ◽  
Nuclear Power ◽  
Reactor Core ◽  
Cooling Water ◽  
Radiation Chemistry ◽  
Operating Conditions ◽  
Water Radiolysis ◽  
Direct Measurements ◽  
Temperature Ranges ◽  
Reactor Operating

TEMPERATURE DEPENDENCE OF PRIMARY SPECIES G(VALUES) FORMED FROM RADIOLYSIS OF WATER BY INTERACTION OF TRITIUM β-PARTICLES. G(values) are important to understand the effect of radiolysis of Nuclear Power Plant (NPP) cooling water. Since direct measurements are difficult, hence modeling and computer simulation were carried out to predict radiation chemistry in and around reactor core. G(values) are required to calculate the radiation chemistry. Monte Carlo simulations were used to calculate the G(values) of primary species , H•, H2, •OH dan H2O2 formed from the radiolysis of tritium β low energy electron. These radiolytic products can degrade the reactor components and cause corrosion under the reactor operating conditions. G(values) prediction can indirectly contribute to maintain the material reliability. G(values) were calculated at 10-8, 10-7, 10-6 and 10-5 s after ionization at temperature ranges. The calculation were compared with the G(values) of g-ray 60Co. The work aimed to understand temperature effect on the water radiolysis mechanism by the tritium β electron. The results show that the trend similarity was found on the temperature dependence of G(values) of tritium β electron and g-ray 60Co. For tritium β electron, G(values) for free radical were lower than g-ray 60Co, but higher for molecular products as temperature raise at 10-8 and 10-7. The significant differences for these two type of radiations were on G(H2), G(•OH) and G(H•) at 10-6and 10-5 s above 200 oC.

scholarly journals CALCULATION OF RADIOACTIVE SOURCE TERM RELEASE FROM FLEXBLUE NPP

2021 ◽  
Vol 23 (2) ◽  
pp. 63
Author(s):  
Muhammad Budi Setiawan ◽  
Pande Made Udiyani
Keyword(s):  
Nuclear Power ◽  
Noble Metals ◽  
Nuclear Reactor ◽  
Source Term ◽  
Reactor Core ◽  
Fission Products ◽  
Operating Conditions ◽  
Normal Operating ◽  
Small Modular Reactor ◽  
Mechanistic Approach

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).

scholarly journals Assessment of Radiation Doses for Lungs due to Some Radionuclides Released from a Hypothetical Nuclear Power Reactor Accident

2015 ◽  
Vol 17 (1) ◽  
pp. 30-37
Author(s):  
Md Moniruzzaman Khan ◽  
AHM Ruhul Quddus ◽  
Mir Md Akramuzzaman ◽  
Abdus Sattar Mollah
Keyword(s):  
Human Body ◽  
Nuclear Power ◽  
Reactor Core ◽  
Reactor Power ◽  
Dose Effect ◽  
Severe Accident ◽  
Reactor Accident ◽  
Nuclear Power Reactor ◽  
North East ◽  
Sensitive Organ

Different radionuclides are emitted from the reactor core after the nuclear accident. These radionuclides are entered into human body through different pathways, which damage the cells. The dose consequence to the sensitive organ like lungs of human body is considered in the present study to show the dose effect for various radionuclides from a hypothetical nuclear reactor accident. The calculations were made with the in-house developed computer program “RaDARRA”. Cardinal directions like E, ENE, ESE, N, NE, NNE, NNW, NW, S, SE, SSE, SSW, SW, W, WNW and WSW are considered to observe the dose effect along the directions. For the calculations, lungs dose arising from 8 radionuclides e.g., 89Sr, 91Y, 95Zr, 95Nb, 131I, 133I, 140Ba and 144Ce have been considered. Of all these radionuclides the maximum and minimum dose contribution mainly come from 144Ce (30%) and 95Nb (4.43%). It is marked that dose is maximum along North East (NE) direction for all the distances and for all types of the radioisotopes. Methodology used in the present study can also be utilized for any type of severe accident and any type of reactor power. DOI: http://dx.doi.org/10.3329/bjnm.v17i1.22489 Bangladesh J. Nuclear Med. 17(1): 30-37, January 2014

scholarly journals Possibility of Simulating Forced Convection in Fast Neutron Reactors Using a Light Water Test Facility

2018 ◽  
Vol 3 (3) ◽  
pp. 279
Author(s):  
V.I. Slobodchuk ◽  
E.A. Avramova ◽  
E.M. Shchennikova ◽  
D.A. Shal'kov
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Similarity Theory ◽  
Test Facility ◽  
Test Model ◽  
Nuclear Power Reactor ◽  
Light Water ◽  
Reactor Installation ◽  
Water Test ◽  
Metal Coolant

The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. A large nuclear power reactor (like the BN-1200 project) was selected as a reactor installation to be modeled. To validate the model, the similarity theory and the “black box” method were used. The paper uses the experience of a number of researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. The governing criteria of similarity were estimated based on the fundamental differential equations of convective heat transfer, so were the conditions under which it is possible to model sodium coolant by using  light water with adequate accuracy. The paper presents the scales of the parameters used for the model - reactor comparison. Dependence curves of certain scales with regard to others are constructed, and the possibility of achieving similarity of certain parameters in modeling was estimated. Recommendations are provided on designing a water test model of the BN reactor and on carrying out experiments using this test model.

Inspection and Evaluation Guidelines for Light Water Reactor Core Internals in Japan

2002 ◽  
Author(s):  
Kimihito Takeuchi ◽  
Naoto Iizuka ◽  
Masashi Kameyama ◽  
Haruo Fujimori ◽  
Yuichi Motora ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Integrity Assessment ◽  
Engineering Society ◽  
Safety Function ◽  
Internal Component

There have been many cracking experiences of light water reactor (LWR) core internals worldwide in the past. Thermal and Nuclear Power Engineering Society in Japan (TENPES) has organized a committee to prepare technically reasonable and appropriate inspection and evaluation guidelines (I&E guidelines) for core internals. This committee consists of scholars and representatives from electric utilities and nuclear plant vendors in Japan. I&E guidelines, which cover a rational inspection plan, structural integrity assessment and repair methods, have been developed considering nuclear safety function and structural strength of each core internal component. For BWR reactors, the development of I&E guidelines cover major core internal components like shroud support, core shroud, top guide, core plate, ICM and CRD housing, core spray piping and sparger, jet pump etc. For PWR reactors, the development of I&E guidelines cover baffle former bolts, barrel former bolts, core barrel weld, bottom mounted instrumentation, etc. The I&E guidelines will be completed by the end of March 2002. The basic concept of the guidelines, and a guideline for shroud support of a BWR as an example, are shown in this paper.

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