Nuclear heating plays an important aspect in design and deployment of both fission and fusion reactors and experimental devices in terms of cooling requirements. Two experimental campaigns in the framework of a collab...Nuclear heating plays an important aspect in design and deployment of both fission and fusion reactors and experimental devices in terms of cooling requirements. Two experimental campaigns in the framework of a collaboration project between the French Atomic and Alternative Energy Commission(CEA) and Jožef Stefan Institute(JSI), Slovenia, have been performed at the JSI TRIGA reactor for the experimental assessment of nuclear heating in fission and fusion-relevant materials by the differential calorimetry technique, based on the CALMOS and CARMEN differential calorimeters, previously developed at CEA. The results of the first campaign performed at reactor powers between 100 and 250 kW have already been reported, highlighting some measurement difficulties. Therefore, the second campaign was performed at a lower reactor power of 30 kW to overcome these issues. Moreover, a computational analysis of the experiments was performed using the JSIR2S code package to calculate the nuclear heating levels. Both experiments and their reproduction by simulations are described in detail. We present a comparison of the previously reported measured nuclear heating values of the first campaign with the computational results, with consistent underestimation by simulations by 8–35%. We report the experimental and computational results for the second experimental campaign performed at a reactor power of 30 kW. The simulated heating values were in agreement with the measurements within the measured heating uncertainty, with simulated heating 2.7–11.3% lower than the experimental values.展开更多
From an engineering feasibility standpoint, what level of performance metrics can be ultimately achieved when designing a reactor using well-established nuclear fuels and structural materials that have already undergo...From an engineering feasibility standpoint, what level of performance metrics can be ultimately achieved when designing a reactor using well-established nuclear fuels and structural materials that have already undergone irradiation testing? The irradiation capability, which hinges on parameters like neutron flux level, irradiation channels' volume, and fuel cycle duration, is a core indicator for high-flux reactors. We propose a conceptual design of an ultra-high flux fast reactor(UFFR) with strong irradiation capability, which utilizes U-20Pu-10Zr alloy fuel and employs lead-bismuth as the coolant. The maximum neutron flux in the core reaches 1.32×10^(16) cm^(-2)s^(-1), while the average neutron flux in the irradiation channels attains 1.19×10^(16) cm^(-2)s^(-1). The volume of the central irradiation channel exceeds 10000 cm^(3), and the fuel cycle duration is 165 d, placing all its performance indicators among the top in the world. Based on the analyses of reactor physics and thermalhydraulics, it has been demonstrated that all reactivity coefficients are negative and all physical parameters meet the design criteria, ensuring the inherent safety of UFFR. An assessment of the irradiation capability has been carried out based on californium-252(^(252)Cf) production, indicating that the irradiation capability of UFFR surpasses that of the high flux isotope reactor(HFIR). The yield of ^(252)Cf from UFFR is 14.39 times that of HFIR, and its nuclei conversion rate is 3.21 times that of HFIR.展开更多
中国绵阳研究堆(China Mianyang Research Reactor,CMRR)是建设于四川省绵阳市的一座多用途高通量研究堆。围绕反应堆中子源高效利用,CMRR现已逐步建成包括材料辐照效应研究、放射性同位素研发、中子科学研究的三大科学平台。CMRR堆采...中国绵阳研究堆(China Mianyang Research Reactor,CMRR)是建设于四川省绵阳市的一座多用途高通量研究堆。围绕反应堆中子源高效利用,CMRR现已逐步建成包括材料辐照效应研究、放射性同位素研发、中子科学研究的三大科学平台。CMRR堆采用常温、常压、深水池淹没设计,确保反应堆在各类事故下的绝对安全。反应堆的倒中子阱结构,在堆芯活性区形成最大快中子注量率3.4×10^(14)n·cm^(-2)·s^(-1),重水箱反射层形成最大热中子注量率2.0×10^(14)n·cm^(-2)·s^(-1),为材料辐照考核、同位素研发及中子科学研究提供充分的实验空间。本文概括性介绍近年来CMRR堆在材料辐照效应研究、放射性同位素研发、中子科学研究的主要研究进展。展开更多
基金supported by the Slovenian Research Agency(research project NC-0001-Analysis of nuclear heating in a reactor,research core funding Reactor physics No.P2-0073,infrastructure program I0-0005)。
文摘Nuclear heating plays an important aspect in design and deployment of both fission and fusion reactors and experimental devices in terms of cooling requirements. Two experimental campaigns in the framework of a collaboration project between the French Atomic and Alternative Energy Commission(CEA) and Jožef Stefan Institute(JSI), Slovenia, have been performed at the JSI TRIGA reactor for the experimental assessment of nuclear heating in fission and fusion-relevant materials by the differential calorimetry technique, based on the CALMOS and CARMEN differential calorimeters, previously developed at CEA. The results of the first campaign performed at reactor powers between 100 and 250 kW have already been reported, highlighting some measurement difficulties. Therefore, the second campaign was performed at a lower reactor power of 30 kW to overcome these issues. Moreover, a computational analysis of the experiments was performed using the JSIR2S code package to calculate the nuclear heating levels. Both experiments and their reproduction by simulations are described in detail. We present a comparison of the previously reported measured nuclear heating values of the first campaign with the computational results, with consistent underestimation by simulations by 8–35%. We report the experimental and computational results for the second experimental campaign performed at a reactor power of 30 kW. The simulated heating values were in agreement with the measurements within the measured heating uncertainty, with simulated heating 2.7–11.3% lower than the experimental values.
基金supported by the National Natural Science Foundation of China (Grant No.12575180)the Lingchuang Research Project of China National Nuclear Corporation (CNNC)。
文摘From an engineering feasibility standpoint, what level of performance metrics can be ultimately achieved when designing a reactor using well-established nuclear fuels and structural materials that have already undergone irradiation testing? The irradiation capability, which hinges on parameters like neutron flux level, irradiation channels' volume, and fuel cycle duration, is a core indicator for high-flux reactors. We propose a conceptual design of an ultra-high flux fast reactor(UFFR) with strong irradiation capability, which utilizes U-20Pu-10Zr alloy fuel and employs lead-bismuth as the coolant. The maximum neutron flux in the core reaches 1.32×10^(16) cm^(-2)s^(-1), while the average neutron flux in the irradiation channels attains 1.19×10^(16) cm^(-2)s^(-1). The volume of the central irradiation channel exceeds 10000 cm^(3), and the fuel cycle duration is 165 d, placing all its performance indicators among the top in the world. Based on the analyses of reactor physics and thermalhydraulics, it has been demonstrated that all reactivity coefficients are negative and all physical parameters meet the design criteria, ensuring the inherent safety of UFFR. An assessment of the irradiation capability has been carried out based on californium-252(^(252)Cf) production, indicating that the irradiation capability of UFFR surpasses that of the high flux isotope reactor(HFIR). The yield of ^(252)Cf from UFFR is 14.39 times that of HFIR, and its nuclei conversion rate is 3.21 times that of HFIR.
文摘中国绵阳研究堆(China Mianyang Research Reactor,CMRR)是建设于四川省绵阳市的一座多用途高通量研究堆。围绕反应堆中子源高效利用,CMRR现已逐步建成包括材料辐照效应研究、放射性同位素研发、中子科学研究的三大科学平台。CMRR堆采用常温、常压、深水池淹没设计,确保反应堆在各类事故下的绝对安全。反应堆的倒中子阱结构,在堆芯活性区形成最大快中子注量率3.4×10^(14)n·cm^(-2)·s^(-1),重水箱反射层形成最大热中子注量率2.0×10^(14)n·cm^(-2)·s^(-1),为材料辐照考核、同位素研发及中子科学研究提供充分的实验空间。本文概括性介绍近年来CMRR堆在材料辐照效应研究、放射性同位素研发、中子科学研究的主要研究进展。
