2024, 45(6): 1-8.
doi: 10.13832/j.jnpe.2024.06.0001
Abstract:
To address the challenge of unknown preset spectra, this paper introduces a two-step spectrum unfolding method that combines the generalized regression neural network (GRNN) and the iterative algorithm. We have independently developed the spectrum unfolding codes for GRNN and iteration and conducted separate and comprehensive validations of the codes. Initially, we utilized activation method data from the Chinese Experimental Fast Reactor (CEFR) to validate the codes. The results indicated that at neutron energies greater than 0.1 MeV, the GRNN results deviated by a maximum of 10.36% from the theoretical spectra. The iterative method’s results deviated by a maximum of 9.15% compared to those obtained using the least squares method. The calculated single nuclear reaction rates showed a maximum relative deviation of 11.71% from the experimental values, indicating good agreement. Furthermore, the GRNN method demonstrated higher accuracy compared to the iterative method without accurate pre-set spectra. Finally, comprehensive validation was performed using Russian boron carbide irradiation data, revealing a maximum deviation of 11.42% in the fast neutron region between the two-step method and the iterative method with pre-set spectra. Therefore, employing a "two-step spectrum unfolding method" to address the challenge of unknown pre-set spectra is feasible, with errors remaining within acceptable limits. The innovative spectrum unfolding method introduced in this paper offers fresh perspectives for the spectrum unfolding of new reactors and offers significant reference value for experiments with unknown pre-set spectra.
To address the challenge of unknown preset spectra, this paper introduces a two-step spectrum unfolding method that combines the generalized regression neural network (GRNN) and the iterative algorithm. We have independently developed the spectrum unfolding codes for GRNN and iteration and conducted separate and comprehensive validations of the codes. Initially, we utilized activation method data from the Chinese Experimental Fast Reactor (CEFR) to validate the codes. The results indicated that at neutron energies greater than 0.1 MeV, the GRNN results deviated by a maximum of 10.36% from the theoretical spectra. The iterative method’s results deviated by a maximum of 9.15% compared to those obtained using the least squares method. The calculated single nuclear reaction rates showed a maximum relative deviation of 11.71% from the experimental values, indicating good agreement. Furthermore, the GRNN method demonstrated higher accuracy compared to the iterative method without accurate pre-set spectra. Finally, comprehensive validation was performed using Russian boron carbide irradiation data, revealing a maximum deviation of 11.42% in the fast neutron region between the two-step method and the iterative method with pre-set spectra. Therefore, employing a "two-step spectrum unfolding method" to address the challenge of unknown pre-set spectra is feasible, with errors remaining within acceptable limits. The innovative spectrum unfolding method introduced in this paper offers fresh perspectives for the spectrum unfolding of new reactors and offers significant reference value for experiments with unknown pre-set spectra.
2024, 45(6): 47-54.
doi: 10.13832/j.jnpe.2024.06.0047
Abstract:
High-temperature and high-velocity impact simulation test is a significant experiment to evaluate the safety of space nuclear power reactor in the accident impact on ground after accidental reentry. In this paper, a finite volume model coupling conduction, convection and radiation is established for the heat loading and high-velocity flight phase of the test, and the thermal response characteristics of the core simulator of the space nuclear reactor in the test are numerically studied, and the effects of loading temperature change rate and diameter-height ratio are analyzed. The results show that during the heat loading phase, the highest temperature and the lowest temperature of the core simulator are located at the junction of the side surface and the bottom surface and at the center of the simulator respectively. The time to reach thermal equilibrium is not only affected by the change rate of loading temperature, but also depends on the diameter-height ratio of the simulator. In the high-velocity flight phase, the highest and lowest temperatures of the core simulator are opposite to those in the heat loading phase, and the lowest temperature decreases with the increase of the diameter-height ratio and flight time. The research results can support the development and experimental design of high-temperature and high-velocity impact simulation test system.
High-temperature and high-velocity impact simulation test is a significant experiment to evaluate the safety of space nuclear power reactor in the accident impact on ground after accidental reentry. In this paper, a finite volume model coupling conduction, convection and radiation is established for the heat loading and high-velocity flight phase of the test, and the thermal response characteristics of the core simulator of the space nuclear reactor in the test are numerically studied, and the effects of loading temperature change rate and diameter-height ratio are analyzed. The results show that during the heat loading phase, the highest temperature and the lowest temperature of the core simulator are located at the junction of the side surface and the bottom surface and at the center of the simulator respectively. The time to reach thermal equilibrium is not only affected by the change rate of loading temperature, but also depends on the diameter-height ratio of the simulator. In the high-velocity flight phase, the highest and lowest temperatures of the core simulator are opposite to those in the heat loading phase, and the lowest temperature decreases with the increase of the diameter-height ratio and flight time. The research results can support the development and experimental design of high-temperature and high-velocity impact simulation test system.
2024, 45(6): 98-105.
doi: 10.13832/j.jnpe.2024.06.0098
Abstract:
In this paper, CPR1000 is taken as the reference unit. According to the Level 1 CPR1000 Probabilistic Safety Analysis (PSA) results, the Large Break LOCA, Intermediate Break LOCA, Small Break LOCA, Station Blackout (SBO), Total Loss of Feedwater (TLOFW) and Anticipated Transient without Trip (ATWT) with Loss of Main Feedwater (LOMF) are selected as the representative design extend condition (DEC) accident scenarios. Using LOCUST and SPRUCE, the thermal and hydraulic codes developed by China Nuclear Power Technology Research Institute Co., Ltd. based on the performance of accident tolerant fuel (ATF), deterministic calculations are carried out for the five types of ATF under development, namely, ATF-1, ATF-2, ATF-3, ATF-4, and ATF-5. Compared with the traditional UO2-Zr material, the accident process, core damage time, system success criteria and personnel response time of different ATFs under the above typical accidents are analyzed. It is found that the lower peak cladding temperature and higher cladding limit temperature of ATF in the accident make CPR1000 unit have greater safety margin, which provides support for ATF material selection. Based on the results of deterministic analysis, the Level 1 PSA model is established for different ATF, and the influence of different ATF materials on the safety of CPR1000 unit is given from the perspective of probability theory. The results show that there is no substantial benefit from the direct application of existing ATF to existing reactors. Based on the deterministic and probabilistic analyses, the development direction of reactor based on ATF is given in this paper.
In this paper, CPR1000 is taken as the reference unit. According to the Level 1 CPR1000 Probabilistic Safety Analysis (PSA) results, the Large Break LOCA, Intermediate Break LOCA, Small Break LOCA, Station Blackout (SBO), Total Loss of Feedwater (TLOFW) and Anticipated Transient without Trip (ATWT) with Loss of Main Feedwater (LOMF) are selected as the representative design extend condition (DEC) accident scenarios. Using LOCUST and SPRUCE, the thermal and hydraulic codes developed by China Nuclear Power Technology Research Institute Co., Ltd. based on the performance of accident tolerant fuel (ATF), deterministic calculations are carried out for the five types of ATF under development, namely, ATF-1, ATF-2, ATF-3, ATF-4, and ATF-5. Compared with the traditional UO2-Zr material, the accident process, core damage time, system success criteria and personnel response time of different ATFs under the above typical accidents are analyzed. It is found that the lower peak cladding temperature and higher cladding limit temperature of ATF in the accident make CPR1000 unit have greater safety margin, which provides support for ATF material selection. Based on the results of deterministic analysis, the Level 1 PSA model is established for different ATF, and the influence of different ATF materials on the safety of CPR1000 unit is given from the perspective of probability theory. The results show that there is no substantial benefit from the direct application of existing ATF to existing reactors. Based on the deterministic and probabilistic analyses, the development direction of reactor based on ATF is given in this paper.
2024, 45(6): 132-138.
doi: 10.13832/j.jnpe.2024.06.0132
Abstract:
Noise in the main control room is one of the major concerns for operational safety of the nuclear power plant. In this paper, the impact of the main steam pipeline vibration on the noise level in the main control room of the high temperature gas-cooled reactor (HTGR) is investigated by using a structural finite element model and an acoustic boundary element model. A finite element model for frequency response analysis of the nuclear island of one HTGR conceptual design, and a boundary element model for acoustic analysis of the main control room in the frequency domain, are established respectively, to predict the noise levels in the main control room dominated by vibration transfer from the main steam pipelines. The influence patterns of various main steam pipes are explored, and the wall vibration that contributes most to the noise level of the main control room is identified based on acoustic contribution analysis. A method for optimizing the main control room noise through physical partitioning is proposed. The results show that, horizontal vibrations of the main steam pipeline generate higher noise levels in the control room compared to vertical vibrations. The maximum noise caused by the vibration of the main steam pipeline exceeds 60 dB. The walls near the main steam isolation valve room and the ceiling contribute the most to the indoor noise in the control room. Through physical partitioning, the noise level of the control room can be reduced significantly.
Noise in the main control room is one of the major concerns for operational safety of the nuclear power plant. In this paper, the impact of the main steam pipeline vibration on the noise level in the main control room of the high temperature gas-cooled reactor (HTGR) is investigated by using a structural finite element model and an acoustic boundary element model. A finite element model for frequency response analysis of the nuclear island of one HTGR conceptual design, and a boundary element model for acoustic analysis of the main control room in the frequency domain, are established respectively, to predict the noise levels in the main control room dominated by vibration transfer from the main steam pipelines. The influence patterns of various main steam pipes are explored, and the wall vibration that contributes most to the noise level of the main control room is identified based on acoustic contribution analysis. A method for optimizing the main control room noise through physical partitioning is proposed. The results show that, horizontal vibrations of the main steam pipeline generate higher noise levels in the control room compared to vertical vibrations. The maximum noise caused by the vibration of the main steam pipeline exceeds 60 dB. The walls near the main steam isolation valve room and the ceiling contribute the most to the indoor noise in the control room. Through physical partitioning, the noise level of the control room can be reduced significantly.
2024, 45(6): 139-146.
doi: 10.13832/j.jnpe.2024.06.0139
Abstract:
Control rod position sensor is one of the six core components of the control rod hydraulic drive system, which provides the only true rod position indication for nuclear reactors. The grounding measurement capacitance rod position sensor has the advantages of high precision and strong anti-interference ability. Control rod positions can be measured step-by-step by this kind of sensors. In order to clarify the capacitive sensitive mechanism of this kind of capacitance rod position sensors, the sensitivity analysis model of the sensor is established by conformal mapping. Model modification and model validation are conducted by the numerical simulation and results of the static calibration experiments. The results show that the static measurement characteristics of the grounding measurement capacitance rod position sensor can be accurately analyzed by the sensitivity analysis model. The relative error between the results obtained by the sensitive analysis model and experiments is 3.4%. The sensitivity analysis model can be used for structural analysis and optimal design of the sensor.
Control rod position sensor is one of the six core components of the control rod hydraulic drive system, which provides the only true rod position indication for nuclear reactors. The grounding measurement capacitance rod position sensor has the advantages of high precision and strong anti-interference ability. Control rod positions can be measured step-by-step by this kind of sensors. In order to clarify the capacitive sensitive mechanism of this kind of capacitance rod position sensors, the sensitivity analysis model of the sensor is established by conformal mapping. Model modification and model validation are conducted by the numerical simulation and results of the static calibration experiments. The results show that the static measurement characteristics of the grounding measurement capacitance rod position sensor can be accurately analyzed by the sensitivity analysis model. The relative error between the results obtained by the sensitive analysis model and experiments is 3.4%. The sensitivity analysis model can be used for structural analysis and optimal design of the sensor.
2024, 45(6): 159-165.
doi: 10.13832/j.jnpe.2024.06.0159
Abstract:
Pulsed neutron detector converts neutron fluence rate into random weak current pulse signal. Due to the particularity of this signal, nuclear measurement equipment generally requires reactor tests to verify the actual detection performance. Because of the research method by reactor test costs a lot and has time limit, this paper, based on a typical pulsed neutron detector such as boron-coated proportional counting tube, studies a pulsed detector signal model and its nuclear signal generator implementation scheme. The characteristics of each key part are verified through simulation. The verification results show that: the proposed nuclear signal generator scheme can generate a sequence of time intervals satisfying the exponential distribution, the single current pulse shape is similar to the detector and the amplitude can vary randomly in a uniform distribution.
Pulsed neutron detector converts neutron fluence rate into random weak current pulse signal. Due to the particularity of this signal, nuclear measurement equipment generally requires reactor tests to verify the actual detection performance. Because of the research method by reactor test costs a lot and has time limit, this paper, based on a typical pulsed neutron detector such as boron-coated proportional counting tube, studies a pulsed detector signal model and its nuclear signal generator implementation scheme. The characteristics of each key part are verified through simulation. The verification results show that: the proposed nuclear signal generator scheme can generate a sequence of time intervals satisfying the exponential distribution, the single current pulse shape is similar to the detector and the amplitude can vary randomly in a uniform distribution.
2024, 45(6): 220-225.
doi: 10.13832/j.jnpe.2024.06.0220
Abstract:
The operation data of PWR nuclear power plant show that after the unit implements large flow degassing operation, the specific activity of fission products of primary coolant oscillates violently in a short time, which makes the fuel damage prediction method based on the average core state fission release-to-birth ratio (R/B) have prediction bias. Based on the parameters of degassing system and the mechanism of inert gas release in PWR nuclear power plant, this paper establishes a modified prediction and analysis model of inert gas release in degassing operation, gives the calculation method of degassing factor and inert gas release rate under degassing conditions, and optimizes the traditional prediction method of fuel rod break size based on R/B. The modified prediction method of degassing operation has been applied and verified in a PWR nuclear power plant. The maximum relative deviation of specific activities of six common inert gas nuclides predicted is 33.4%, and the others are less than 20%. The predicted fuel rod break size is large, which is consistent with the inspection results after shutdown.
The operation data of PWR nuclear power plant show that after the unit implements large flow degassing operation, the specific activity of fission products of primary coolant oscillates violently in a short time, which makes the fuel damage prediction method based on the average core state fission release-to-birth ratio (R/B) have prediction bias. Based on the parameters of degassing system and the mechanism of inert gas release in PWR nuclear power plant, this paper establishes a modified prediction and analysis model of inert gas release in degassing operation, gives the calculation method of degassing factor and inert gas release rate under degassing conditions, and optimizes the traditional prediction method of fuel rod break size based on R/B. The modified prediction method of degassing operation has been applied and verified in a PWR nuclear power plant. The maximum relative deviation of specific activities of six common inert gas nuclides predicted is 33.4%, and the others are less than 20%. The predicted fuel rod break size is large, which is consistent with the inspection results after shutdown.
2024, 45(6): 248-253.
doi: 10.13832/j.jnpe.2024.06.0248
Abstract:
The development of advanced reactor designs imposes higher demands on neutronic numerical methods. To achieve accurate and efficient simulation of complex problems, this paper introduces a hybrid discontinuous Galerkin (HDG) method based on the first-order hyperbolic neutron transport equation (NTE). The method decouples the original equation into independent equations for each angular direction using the discrete-ordinates (SN) method in angular space. In spatial discretization, this paper employs an upwind scheme that results in a blocked-lower-triangular global matrix coupling system, making it well-suited for complex, geometrically heterogeneous neutron transport scenarios with a large number of meshes. The study evaluates the performance of the proposed HDG method using the TAKEDA1 benchmark and a heterogeneous assembly problem. The results demonstrate that the HDG method achieves stable convergence for the aforementioned problems, with a maximum error between the effective multiplication coefficient keff and the reference solution of 108 pcm (1pcm = 10−5). In addition, compared with the traditional second-order even-parity method, the first-order HDG method is more efficient in spatial scanning, and the acceleration ratio is about 2 times in the above examples. Therefore, the proposed HDG method can provide an alternative solution for complex reactor problems.
The development of advanced reactor designs imposes higher demands on neutronic numerical methods. To achieve accurate and efficient simulation of complex problems, this paper introduces a hybrid discontinuous Galerkin (HDG) method based on the first-order hyperbolic neutron transport equation (NTE). The method decouples the original equation into independent equations for each angular direction using the discrete-ordinates (SN) method in angular space. In spatial discretization, this paper employs an upwind scheme that results in a blocked-lower-triangular global matrix coupling system, making it well-suited for complex, geometrically heterogeneous neutron transport scenarios with a large number of meshes. The study evaluates the performance of the proposed HDG method using the TAKEDA1 benchmark and a heterogeneous assembly problem. The results demonstrate that the HDG method achieves stable convergence for the aforementioned problems, with a maximum error between the effective multiplication coefficient keff and the reference solution of 108 pcm (1pcm = 10−5). In addition, compared with the traditional second-order even-parity method, the first-order HDG method is more efficient in spatial scanning, and the acceleration ratio is about 2 times in the above examples. Therefore, the proposed HDG method can provide an alternative solution for complex reactor problems.
2024, 45(6): 263-270.
doi: 10.13832/j.jnpe.2024.06.0263
Abstract:
Aiming at the compatibility problem of high-temperature flowing liquid metal on structural materials, especially the corrosion problem, in the first wall of liquid lithium and liquid metal blanket components in the nuclear fusion reactor, a high-temperature flowing liquid metal corrosion experimental device is designed, and three-dimensional numerical simulation and analysis of the flow and heat transfer characteristics of the liquid metal are carried out by using the software ANSYS. The simulation and test results show that the experimental device can realize the conditions of liquid lithium temperature (300-600℃) and flow rate (< 0.2 m/s) in the first wall and blanket structure, and is qualified to study the corrosion characteristics of dynamic liquid lithium and structural materials at high temperature. Meanwhile, the corrosion behavior of domestically produced low-activation ferrite/martensite steel (9Cr-0.4Mo-0.3Y steel) in 0.2 m/s liquid Li at 550℃ for 1000 hours (h) is preliminarily studied. The results show that 9Cr-0.4Mo-0.3Y steel experiences obvious intergranular corrosion and pitting corrosion, and the surface hardness of the sample is reduced to different degrees due to non-uniform corrosion. The XRD analysis reveals that there is no phase transformation on the corroded surface of 9Cr-0.4Mo-0.3Y steel. The 03-1049#FeNi peak is detected on the sample' surface due to the dissolution and migration of Ni element from the 304 stainless steel vessel.
Aiming at the compatibility problem of high-temperature flowing liquid metal on structural materials, especially the corrosion problem, in the first wall of liquid lithium and liquid metal blanket components in the nuclear fusion reactor, a high-temperature flowing liquid metal corrosion experimental device is designed, and three-dimensional numerical simulation and analysis of the flow and heat transfer characteristics of the liquid metal are carried out by using the software ANSYS. The simulation and test results show that the experimental device can realize the conditions of liquid lithium temperature (300-600℃) and flow rate (< 0.2 m/s) in the first wall and blanket structure, and is qualified to study the corrosion characteristics of dynamic liquid lithium and structural materials at high temperature. Meanwhile, the corrosion behavior of domestically produced low-activation ferrite/martensite steel (9Cr-0.4Mo-0.3Y steel) in 0.2 m/s liquid Li at 550℃ for 1000 hours (h) is preliminarily studied. The results show that 9Cr-0.4Mo-0.3Y steel experiences obvious intergranular corrosion and pitting corrosion, and the surface hardness of the sample is reduced to different degrees due to non-uniform corrosion. The XRD analysis reveals that there is no phase transformation on the corroded surface of 9Cr-0.4Mo-0.3Y steel. The 03-1049#FeNi peak is detected on the sample' surface due to the dissolution and migration of Ni element from the 304 stainless steel vessel.
2018, 39(5): 1-3.
doi: 10.13832/j.jnpe.2018.05.0001
摘要：
主要介绍了我国在建、在运核电机组的基本状况和最新进展，以及我国在提升核设施安全水平方面的相关措施。在国家能源局印发的《能源技术创新“十三五”规划》要求之下，我国推出一系列先进核能和小型堆的发展计划，开展了“海洋核动力平台示范工程建设”并建立相关标准。最后总结了中国核电目前面临的挑战和未来的展望。
主要介绍了我国在建、在运核电机组的基本状况和最新进展，以及我国在提升核设施安全水平方面的相关措施。在国家能源局印发的《能源技术创新“十三五”规划》要求之下，我国推出一系列先进核能和小型堆的发展计划，开展了“海洋核动力平台示范工程建设”并建立相关标准。最后总结了中国核电目前面临的挑战和未来的展望。
2019, 40(4): 1-8.
摘要：
热管冷却反应堆采用固态反应堆设计理念，通过热管非能动方式导出堆芯热量。本文总结了热管冷却反应堆的概念初创、积极探索、重大突破的发展历程；分析了热管冷却反应堆的技术特点，包括固态属性、固有安全性高、运行特性简单、易于模块化与易扩展和运输特性良好等核心优势；归纳了热管冷却反应堆中热管性能、材料工艺、能量转换等技术现状，并提出热管冷却反应堆进一步发展将面临的材料、制造工艺、运行可维护性等挑战，从而明确了热管冷却反应堆未来的发展趋势，为革新型热管冷却反应堆技术的发展与应用提供良好的方向指引。总体而言，热管冷却反应堆在深空探测与推进、陆基核电源、深海潜航探索等场景中具有广阔的应用前景，有可能成为改变未来核动力格局的颠覆性技术之一。
热管冷却反应堆采用固态反应堆设计理念，通过热管非能动方式导出堆芯热量。本文总结了热管冷却反应堆的概念初创、积极探索、重大突破的发展历程；分析了热管冷却反应堆的技术特点，包括固态属性、固有安全性高、运行特性简单、易于模块化与易扩展和运输特性良好等核心优势；归纳了热管冷却反应堆中热管性能、材料工艺、能量转换等技术现状，并提出热管冷却反应堆进一步发展将面临的材料、制造工艺、运行可维护性等挑战，从而明确了热管冷却反应堆未来的发展趋势，为革新型热管冷却反应堆技术的发展与应用提供良好的方向指引。总体而言，热管冷却反应堆在深空探测与推进、陆基核电源、深海潜航探索等场景中具有广阔的应用前景，有可能成为改变未来核动力格局的颠覆性技术之一。
2019, 40(6): 1-6.
doi: 10.13832/j.jnpe.2019.06.0001
摘要：
放射性废液得到有效处理是世界各国核工业迅猛发展的前提，其关键技术的现状和发展方向也是我国核工业界关注的焦点。本文介绍了几种放射性废液处理的传统方法及涌现出的新技术，概述了各种方法的原理及优、缺点，同时讨论了放射性废液处理技术今后的研究方向及发展趋势。
放射性废液得到有效处理是世界各国核工业迅猛发展的前提，其关键技术的现状和发展方向也是我国核工业界关注的焦点。本文介绍了几种放射性废液处理的传统方法及涌现出的新技术，概述了各种方法的原理及优、缺点，同时讨论了放射性废液处理技术今后的研究方向及发展趋势。
2019, 40(3): 103-108.
doi: 10.13832/j.jnpe.2019.03.0103
摘要：
以配置四取中逻辑输入模块的核电厂稳压器数字压力控制装置为研究对象，建立其故障树模型，包括四取中逻辑的动态部分和其他设备的静态部分，采用马尔科夫方法分析动态部分，再根据逻辑关系分析整体故障树，最后，围绕可靠度和重要度评价四取中逻辑的可靠性及其对整个装置可靠性的提升效果，结果表明：四取中逻辑在可靠性方面优化程度相对较高。
以配置四取中逻辑输入模块的核电厂稳压器数字压力控制装置为研究对象，建立其故障树模型，包括四取中逻辑的动态部分和其他设备的静态部分，采用马尔科夫方法分析动态部分，再根据逻辑关系分析整体故障树，最后，围绕可靠度和重要度评价四取中逻辑的可靠性及其对整个装置可靠性的提升效果，结果表明：四取中逻辑在可靠性方面优化程度相对较高。
2019, 40(1): 1-7.
doi: 10.13832/j.jnpe.2019.01.0001
摘要：
“华龙一号”是我国自主设计研发的具有完整知识产权的第三代百万千瓦级压水堆核电技术。本文介绍了“华龙一号”的产生历程，系统论述了“华龙一号”反应堆堆芯与安全设计特点，包括“华龙一号”研发过程中开展的堆芯核设计、热工水力设计、安全设计、设计验证及“华龙一号”持续开展的设计改进与优化等内容，通过采用新的设计理念和设计技术，全面提高了“华龙一号”作为三代核电技术的经济性、灵活性和安全性。
“华龙一号”是我国自主设计研发的具有完整知识产权的第三代百万千瓦级压水堆核电技术。本文介绍了“华龙一号”的产生历程，系统论述了“华龙一号”反应堆堆芯与安全设计特点，包括“华龙一号”研发过程中开展的堆芯核设计、热工水力设计、安全设计、设计验证及“华龙一号”持续开展的设计改进与优化等内容，通过采用新的设计理念和设计技术，全面提高了“华龙一号”作为三代核电技术的经济性、灵活性和安全性。
2020, 41(5): 132-137.
摘要：
为解决核电厂传统监测手段的局限性，提出将核主元分析法（KPCA）引入核电厂设备在线监测领域中，并设计了监测模型建设方法以及在线监测策略。为验证算法的有效性，将其应用在国内某核电机组电动主给水泵的真实监测案例中。仿真结果表明，KPCA算法可适应核电厂设备监测的要求，能比现有阈值监测手段提供更为早期的故障预警。同时，相比于常规的主元分析法（PCA），KPCA算法能够提取各变量之间的非线性关系，识别出设备不同的运行模式，有效减少误报警。
为解决核电厂传统监测手段的局限性，提出将核主元分析法（KPCA）引入核电厂设备在线监测领域中，并设计了监测模型建设方法以及在线监测策略。为验证算法的有效性，将其应用在国内某核电机组电动主给水泵的真实监测案例中。仿真结果表明，KPCA算法可适应核电厂设备监测的要求，能比现有阈值监测手段提供更为早期的故障预警。同时，相比于常规的主元分析法（PCA），KPCA算法能够提取各变量之间的非线性关系，识别出设备不同的运行模式，有效减少误报警。
2019, 40(1): 74-77.
doi: 10.13832/j.jnpe.2019.01.0074
摘要：
介绍了中广核研究院在事故容错燃料（ATF）包壳领域的最新成果，通过预置粉末式脉冲激光熔覆技术，在不同的功率下制备出不同厚度的锆包壳管Cr保护层；通过高温蒸汽氧化增重数据发现，采用半导体脉冲激光熔覆技术、脉冲激光功率50~60 W、螺距0.8~0.9 mm、角速度10°/s等参数条件下制备Cr涂层可以获得较好的抗高温氧化性能，证明保护的效果直接受涂层质量控制。通过SEM分析了涂层的显微结构，采用扩散机理解释了Cr涂层在1200℃下与锆合金基体相容性良好的原因。
介绍了中广核研究院在事故容错燃料（ATF）包壳领域的最新成果，通过预置粉末式脉冲激光熔覆技术，在不同的功率下制备出不同厚度的锆包壳管Cr保护层；通过高温蒸汽氧化增重数据发现，采用半导体脉冲激光熔覆技术、脉冲激光功率50~60 W、螺距0.8~0.9 mm、角速度10°/s等参数条件下制备Cr涂层可以获得较好的抗高温氧化性能，证明保护的效果直接受涂层质量控制。通过SEM分析了涂层的显微结构，采用扩散机理解释了Cr涂层在1200℃下与锆合金基体相容性良好的原因。
2020, 41(3): 221-226.
doi: 10.13832/j.jnpe.2020.03.0221
摘要：
为提高已投入运行核动力装置旋转设备的运行数据采集和状态监测能力，需要解决安装传感器和敷设配套线缆困难的问题。本文采用现场可编程门阵列（FPGA）作为主控单元，设计了一种基于Zigbee物联网通信技术的智能无线振动传感器，并给出了其电路构成、工作原理，以及嵌入式控制软件的工作流程。通过对此传感器进行性能测试，结果表明该传感器功耗低，实现了对振动信号的连续采集、智能分析与上传。该无线传感器安装简单，无需敷设供电和信号线缆，可应用于构建核动力装置旋转设备的状态监测系统。
为提高已投入运行核动力装置旋转设备的运行数据采集和状态监测能力，需要解决安装传感器和敷设配套线缆困难的问题。本文采用现场可编程门阵列（FPGA）作为主控单元，设计了一种基于Zigbee物联网通信技术的智能无线振动传感器，并给出了其电路构成、工作原理，以及嵌入式控制软件的工作流程。通过对此传感器进行性能测试，结果表明该传感器功耗低，实现了对振动信号的连续采集、智能分析与上传。该无线传感器安装简单，无需敷设供电和信号线缆，可应用于构建核动力装置旋转设备的状态监测系统。
2020, 41(3): 137-142.
doi: 10.13832/j.jnpe.2020.03.0137
摘要：
为分析核电厂应急人员在处理严重事故时可能发生的人因失误，通过建立不同应急人员的认知模型及识别相应的行为影响因子，在认知功能的基础上识别出13种人因失误模式：信息来源不足、信息可靠性不佳、过早结束对参数的获取、重要数据处理不正确、缓解措施负面影响评估失误、选择不适用当前情景的策略、延迟决策、遗漏重要信息/警报、延迟发觉、软操作失误、信息反馈失效、设备安装/连接/操作失误、延迟实施，并基于故障树分析得出人因失误模式的主要根原因：交流失效、时间压力、事故发展的不确定性、信息接收延误、监视失误、人-机界面不佳和环境因素。分析结果可用于预测严重事故缓解进程中可能出现的人因失误，为核电厂实施严重事故管理和技术改进，以及保障严重事故工况下核电厂安全提供参考。
为分析核电厂应急人员在处理严重事故时可能发生的人因失误，通过建立不同应急人员的认知模型及识别相应的行为影响因子，在认知功能的基础上识别出13种人因失误模式：信息来源不足、信息可靠性不佳、过早结束对参数的获取、重要数据处理不正确、缓解措施负面影响评估失误、选择不适用当前情景的策略、延迟决策、遗漏重要信息/警报、延迟发觉、软操作失误、信息反馈失效、设备安装/连接/操作失误、延迟实施，并基于故障树分析得出人因失误模式的主要根原因：交流失效、时间压力、事故发展的不确定性、信息接收延误、监视失误、人-机界面不佳和环境因素。分析结果可用于预测严重事故缓解进程中可能出现的人因失误，为核电厂实施严重事故管理和技术改进，以及保障严重事故工况下核电厂安全提供参考。
2020, 41(4): 191-196.
摘要：
热管冷却反应堆（简称“热管堆”）高温运行下的结构热膨胀效应会显著影响反应堆的传热和中子物理输运过程。本文提出了一种考虑固体堆芯显著膨胀的几何更新和反应性反馈方法，并构建了基于动态几何的中子物理/热工/力学3场核热力耦合分析程序。在核热力耦合中主要考虑温度引起微观截面的变化、材料密度的变化以及热膨胀引起堆芯尺寸的变化。基于提出的核热力耦合方法，对MegaPower热管堆进行了核热力耦合分析，分析了不同松弛因子下，堆芯功率分布和径向功率因子的收敛性。核热力计算表明，热膨胀造成堆芯边通道的中子泄漏增加，从而产生负反应性反馈；同时，边通道中子泄漏增加加剧了功率分布的不均匀性，传热恶化，考虑核热力耦合后，径向功率因子从非耦合情形的1.20提升到1.23，燃料峰值温度增加11 K。
热管冷却反应堆（简称“热管堆”）高温运行下的结构热膨胀效应会显著影响反应堆的传热和中子物理输运过程。本文提出了一种考虑固体堆芯显著膨胀的几何更新和反应性反馈方法，并构建了基于动态几何的中子物理/热工/力学3场核热力耦合分析程序。在核热力耦合中主要考虑温度引起微观截面的变化、材料密度的变化以及热膨胀引起堆芯尺寸的变化。基于提出的核热力耦合方法，对MegaPower热管堆进行了核热力耦合分析，分析了不同松弛因子下，堆芯功率分布和径向功率因子的收敛性。核热力计算表明，热膨胀造成堆芯边通道的中子泄漏增加，从而产生负反应性反馈；同时，边通道中子泄漏增加加剧了功率分布的不均匀性，传热恶化，考虑核热力耦合后，径向功率因子从非耦合情形的1.20提升到1.23，燃料峰值温度增加11 K。
2022, 43(1): 1-6.
doi: 10.13832/j.jnpe.2022.01.0001
Abstract:
2020, 41(4): 1-7.
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2023, 44(1): 1-8.
doi: 10.13832/j.jnpe.2023.01.0001
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2021, 42(4): 1-5.
doi: 10.13832/j.jnpe.2021.04.0001
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2019, 40(6): 1-6.
doi: 10.13832/j.jnpe.2019.06.0001
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2022, 43(4): 1-4.
doi: 10.13832/j.jnpe.2022.04.0001
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2019, 40(4): 1-8.
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2022, 43(2): 1-8.
doi: 10.13832/j.jnpe.2022.02.0001
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2018, 39(3): 6-9.
doi: 10.13832/j.jnpe.2018.03.0006
Abstract:
2019, 40(5): 1-5.
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