共查询到20条相似文献,搜索用时 31 毫秒
1.
《Progress in Nuclear Energy》2012,54(8):1197-1203
Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. 相似文献
2.
Antonella L. Costa Patrícia Amélia L. Reis Cláubia Pereira Maria Auxiliadora F. Veloso Amir Z. Mesquita 《Nuclear Engineering and Design》2010,240(6):1487-1494
The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data. 相似文献
3.
Patrícia A.L. Reis Antonella L. Costa Cláubia Pereira Maria A.F. Veloso Amir Z. Mesquita Humberto V. Soares Graiciany de P. Barros 《Annals of Nuclear Energy》2010
RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations. 相似文献
4.
Rose Mary Gomes do Prado Souza Amir Zacarias Mesquita 《Progress in Nuclear Energy》2011,53(8):1132-1135
Considering that the power of the IPR-R1 TRIGA reactor, located at the Nuclear Technology Development Center, Brazil, will be increased from 100 kW to 250 kW, some experiments were done in order to evaluate the magnitude of the reactivity effects associated with the reactor operation. The core excess of reactivity obtained was 1.99 $, and the shutdown margin was 1.33 $. The reactivity needed to operate the IPR-R1 reactor at 100 kW was 0.72 $, mainly due to the prompt negative temperature coefficient. A significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels. The loss of reactivity due to xenon poisoning after 8 h of operation at 100 kW was around 0.20 $, and the highest reactivity loss value caused by a void inserted in the central thimble was 0.22 $. From the results obtained, it was possible to balance all the determined reactivity losses with the reactivity excess available in the reactor, considering the present and the future reactor power operation. 相似文献
5.
Patrícia A.L. Reis Antonella L. Costa Claubia Pereira Clarysson A.M. Silva Maria Auxiliadora F. Veloso Amir Z. Mesquita 《Nuclear Engineering and Design》2012
The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system. 相似文献
6.
Rose Mary Gomes do Prado Souza Amir Zacarias Mesquita 《Progress in Nuclear Energy》2011,53(8):1126-1131
The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center - CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state. 相似文献
7.
Today's environmental concerns show that nuclear energy is an important option for meeting future increases in global energy demand. Significant nuclear expansion will probably require new reactor designs in which safety is ensured by simple, convincing means. PIUS represents such a reactor design. It is a re-configured 600 MWe PWR, in which the primary safety goal, protection of the reactor core integrity, is entrusted to built-in, self-protective, passive features, without reliance on any monitoring, detection or actuation system, nor operator action. Its basic design features a core that is openly connected, in a natural circulation loop, to a large pool of borated water. The pool is enclosed in a prestressed concrete pressure vessel provided with redundant leakage barriers. The reactor coolant pumps are operated in such a way that there is hydraulic balance in the openings between the primary coolant loop and the pool. Thereby, the hot, low boron content primary loop water is kept separated from the pool water, in spite of the always open natural circulation loop. In severe transients this balance is disturbed, and pool water ingress occurs, shutting down the reactor, or restricting the power to a safe level. The decay heat is transferred to the pool by the natural circulation loop, and a passive pool cooling system, utilizing natural circulation and natural draft cooling towers, prevents boiling of the pool water, even in a station blackout situation. Transient analyses have shown that this passive long-term RHR function will be available in all accident situations, even after double-ended cold leg breaks. Such breaks result in a temporary pressurization of the reactor containment, but the releases of radioactivity will be extremely small and the doses at the fence boundary very low. Cost estimates indicate that PIUS will be quite competitive, and evaluation studies are now under way in several countries. 相似文献
8.
Maria A. Fortini Michel N. StamoulisÂngela F.M. Ferreira Claubia PereiraAntonella L. Costa Clarysson A.M. Silva 《Annals of Nuclear Energy》2008
In this work, an analytical model for the determination of the temperature distribution in cylindrical heater components with characteristics of nuclear fuel rods, is presented. The heat conductor is characterized by an arbitrary number of solid walls and different types of materials, whose thermal properties are taken as function of temperature. The heat conduction fundamental equation is solved numerically with the method of weighted residuals (MWR) using a technique of orthogonal collocation. The results obtained with the proposed method are compared with the experimental data from tests performed in the TRIGA IPR-R1 research reactor localized at the CDTN/CNEN (Centro de Desenvolvimento da Tecnologia Nuclear/Comissão Nacional de Energia Nuclear) at Belo Horizonte in Brazil. 相似文献
9.
《Progress in Nuclear Energy》2012,54(8):1126-1131
The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state. 相似文献
10.
核电站乏燃料贮存水池失去最终热阱时的安全分析 总被引:1,自引:0,他引:1
压水堆核电站一回路和乏燃料贮存水池的设备冷却水由海水冷却器提供.本文假设事故工况下,海水冷却器突然停止工作,利用热平衡方程,计算并分析了乏燃料贮存水池运行的安全性及作为冷却水源冷却其它一回路重要用户的可能性.计算表明:在本文的各种工况下,乏燃料贮存水池运行是安全的;除一种工况外,硼水还具有冷却其它设备的能力. 相似文献
11.
A. Alberman M. Roche P. Couffin S. Bendotti D. J. Moulin J. L. Boutfroy 《Nuclear Engineering and Design》1997,168(1-3)
The requirements set by the operators of the French nuclear programme, chiefly EDF, Framatome and the safety authorities, for the determination of the technological limit of pressurized water reactor fuels in transient conditions, led the CEA to start a major programme for the qualification of experimental irradiations, and particularly power ramps. The ISABELLE 1 loop, installed in the OSIRIS reactor at Saclay, has been equipped since 1992 with sophisticated instrumentation for power determinations by heat balance. A programme of intercomparison with nuclear measurements (γ spectrometry, dosimetry, radiochemical analysis) helped to qualify these thermal measurements. Thermohydraulic and neutron modelling of the ISABELLE 1 loop substantiated the various corrections to be made to the heat balance. The reproducibility of the test was established by using a neutron signal to drive the position of the loop according to the ramp sequence, and by a special quality assurance programme. Cladding failures by pellet-cladding interaction were systematically observed in the maximum power zone determined by γ spectrometry. 相似文献
12.
Clarysson A.M. da Silva Claubia Pereira Bruno T. Guerra Maria Auxiliadora F. Veloso Antonella L. Costa Hugo M. Dalle Maria Ângela de B.C. Menezes 《Nuclear Engineering and Design》2011,241(12):4989-4993
The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10?4. Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition. 相似文献
13.
14.
Mohamed S. El-Genk 《Nuclear Engineering and Design》2008,238(9):2245-2255
Nuclear reactor power systems could revolutionize space exploration and support human outpost on the moon and Mars. This paper reviews various energy conversion technologies for use in space reactor power systems and provides estimates of the system's net efficiency and specific power, and the specific area of the radiator. The suitable combinations of the energy conversion technologies and the nuclear reactors, classified based on the coolant type and cooling method, for best system performance and highest specific power, are also discussed. In addition, a number of power system concepts with both static and dynamic energy conversion, but with no single point failures in reactor cooling, energy conversion and heat rejection, and for nominal electrical powers up to 110 kWe, are presented. The first two power systems employ reactors cooled with lithium and sodium heat pipes, SiGe thermoelectric (TE) and alkali-metal thermal-to-electric conversion (AMTEC), and potassium heat pipes radiators. The reactors heat pipes operate at a fraction of the prevailing capillary or sonic limit, and in the case of a multiple heat pipes failure, those in the adjacent modules remove the additional heat load, thus maintaining the reactor adequately cooled and the power system operating at a reduced power. The third power system employs SiGe TE converters and a liquid metal cooled reactor with a divided core into six sectors that are neurotically and thermally coupled, but hydraulically decoupled. Each sector has a separate energy conversion loop, a heat rejection loop, and a rubidium heat pipes radiator panel. When a core sector experiences a loss-of-coolant, the fission power of the reactor is reduced, and that generated in the sector in question is removed by the circulating coolant in the adjacent sectors. The fourth power system employs a gas cooled reactor with a core divided into three identical sectors, and each sector is coupled to a separate Closed Brayton Cycle (CBC) loop with He-Xe binary mixture (40 g/mol) working fluid, a secondary loop with circulating liquid Nak-78, and two water heat pipes radiator panels. 相似文献
15.
针对目前航天技术发展对动力提出的要求,参考国外提出的空间核动力系统设计,提出了新型兆瓦级空间热管反应堆核动力系统概念设计。堆芯为金属锂热管冷却、石墨慢化热中子反应堆,采用转鼓控制反应性,堆芯热量通过热管导出。与国外热管反应堆设计方案中燃料棒与热管相间布置方案不同,本文采用了热管-燃料复合元件,即燃料包裹于热管外壁面。能量转换采用以氦氙混合气体为工质的布雷顿动态热电转换。系统废热通过钠钾合金冷却回路传递到钾热管辐射板,通过辐射换热释放入太空。对热管反应堆堆芯物理及热工进行了初步分析,并对热管辐射板进行了性能分析,结果表明,所设计热管反应堆堆芯在设计功率下满足相应安全性要求,同时热管辐射板具有足够的能力将系统废热导出。 相似文献
16.
为研究一体化布置的核供热堆在发生破口失水事故中破口大小和从中间回路排出热量减少对断流过程的影响,选用不同的破口尺寸和不同的二回路工作状态,在5MW核供热堆热工水力模拟回路HRTL-5上进行了实验研究。稳态运行工况的系统压力为1.5MPa,在发生小破口失水事故后,加热功率维持为额定功率的5%以模拟剩余发热情况。实验研究并比较了不同条件下压力、温度、循环流量、液位和失水量等重要参数的变化。这些实验数据为核供热堆的安全分析提供了实验依据。 相似文献
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18.
冷凝器是核动力装置二回路系统中的重要设备,它的重量和尺寸是影响核动力装置重量、体积及布置等的重要因素。本文利用传热经验关系式和冷凝器工业性试验结果,建立了冷凝器数学模型,该模型包括热平衡计算、阻力计算、振动校核和冷凝器重量、体积计算,编制了相应的程序来验证模型的精确性,并对冷凝器重量、体积受冷却管外径、节距和冷却水流速影响的敏感性进行了分析。利用改进遗传算法对冷凝器重量、体积进行优化设计,结果显示,与原方案相比,采用优化方案后冷凝器重量减小了6.926%,体积减小了12.587%,优化效果显著。 相似文献
19.
For a modular reactor of 200 MW thermal output an inactive after heat removal system has been designed. It consists of a prestressed cast iron pressure vessel with the surrounding reactor cell. Integrated in the cast iron profiles of the reactor cell is a redundant water cooling system based on natural convection. Air cooling towers are provided to cool the water down to ambient temperature. The cooling system covers a wide range of possible wall temperatures without significant changes in water temperature. The structures of the reactor pressure vessel and the cell, their assembly and some results of the engineering work done up to now are described in this paper. 相似文献
20.
The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core. 相似文献