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Operating experience and ways to improve the performance of the service water supply system at the Novovoronezh NPP II (Units 1 and 2) 新沃罗涅日核电站二期(1、2号机组)供水系统运行经验及改进方法
Pub Date : 2020-11-20 DOI: 10.3897/nucet.6.60461
V. Povarov, D. B. Statsura, Dmitry Y. Usachev
The operating experience of Novovoronezh NPP II-1 shows that, in the summer period, the temperature of the cooling water exceeds the design value: this indicates the insufficient performance of the service water supply system. The main factor that has a negative impact on the performance of this system is the formation of carbonate deposits on the cooling tower filler. At Novovoronezh NPP II-1, the cooling tower water distribution system was cleaned from carbonate deposits by the method of combined vibration and aerohydraulic impact. The tested method of cleaning the filler cannot be considered optimal, since the main stage that determines the entire cleaning duration is the assembly/disassembly of the cooling tower filler. It is necessary to continue research on the choice of a strategy for controlling the carbonate deposition rate, taking into account the revealed influence of the design features of the main cooling water pipelines and pipelines of the cooling tower water distribution system on the mechanism of deposit formation in the peripheral spraying area. As compensating measures to ensure the required temperature regime of the turbine plant equipment at Novovoronezh NPP II-1, it is practiced during the summer period to put the standby heat exchangers of the lubrication system and the standby pump of the nonessential services cooling water system into parallel operation. This solution is fraught with the risk of an unplanned decrease in the electrical load if this equipment is turned off in the event of a malfunction. To increase the operating stability of Novovoronezh NPP II-1 and -2 in the summer period, it is proposed to carry out a number of measures aimed at mitigating the negative consequences caused by the elevated service water temperature. Equipment upgrade options are evaluated, e.g., by installing an additional pump for the turbine building services cooling system and (or) laying an additional pipeline to supply part of the makeup water from the Don River directly to the suction pipelines of the pumps of the turbine building services cooling system.
新沃罗涅日二期1号核电站的运行经验表明,在夏季,冷却水温度超过设计值,这表明服务供水系统的性能不足。影响该系统性能的主要因素是在冷却塔填料上形成碳酸盐沉积物。在Novovoronezh核电站II-1,冷却塔配水系统通过振动和气动冲击相结合的方法清除了碳酸盐沉积物。所测试的清洗填料的方法不能被认为是最佳的,因为决定整个清洗持续时间的主要阶段是冷却塔填料的组装/拆卸。考虑到主冷却水管道和冷却塔配水系统管道的设计特点对外围喷涂区沉积形成机理的揭示影响,有必要继续研究控制碳酸盐沉积速率的策略选择。为了保证新沃罗涅日核电站II-1机组汽轮机厂设备所需的温度状态,在夏季实行了润滑系统备用热交换器和非必要服务冷却水系统备用泵并联运行的补偿措施。如果在发生故障时关闭该设备,则该解决方案充满了电气负载意外减少的风险。为了提高新沃罗涅日核电站II-1和2夏季的运行稳定性,建议采取一系列措施,以减轻服务水温升高造成的负面后果。评估设备升级方案,例如,通过安装涡轮建筑服务冷却系统的额外泵和(或)铺设额外的管道,将顿河的部分补给水直接供应给涡轮建筑服务冷却系统的泵的吸入管道。
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引用次数: 0
Multi-criteria analysis of the efficiency of scenarios for the development of the Russian nuclear industry in view of the uncertain prospects for the future 考虑到不确定的未来前景,对俄罗斯核工业发展方案的效率进行多标准分析
Pub Date : 2020-11-20 DOI: 10.3897/nucet.6.60557
A. Zrodnikov, V. Korobeynikov, A. Moseev, A. Egorov
Multi-criteria analysis is used in many areas of research where it is required to compare several alternatives according to a selected set of criteria. Of particular interest is the application of this method for a comparative assessment of the efficiency of scenarios for the development of innovative nuclear systems. The article proposes an approach to the computational substantiation of the step-by-step transfer of the Russian nuclear industry to a two-component nuclear energy system (NES) with a centralized closed nuclear fuel cycle (NFC) based on the multi-criteria analysis method. At the same time, consideration is given to options for the development of the domestic nuclear industry in view of the uncertain prospects for the future. Taking into account various trends in the nuclear energy development, the authors identify the following three groups of possible scenarios. The first group includes ‘growing’ scenarios in which the number of units and their total installed capacity grow over time. The second group assumes that after a certain time of growth of the installed capacities, the stationary level will be reached, in which there will be no time-dependent capacity changes. The third group simulates a decrease in the installed nuclear energy capacities in the country after some growth. To select the most preferable ways of technological development and assess the efficiency of a nuclear energy system, a limited set of selection criteria and performance indicators are used, covering the economy, export potential, competitiveness, efficient SNF and RW management, natural uranium consumption, and innovative development potential. An important part of this work was a detailed analysis of the uncertainties in the weights and input data used to derive the criteria.
多标准分析用于许多研究领域,需要根据一组选定的标准比较几种备选方案。特别令人感兴趣的是应用这种方法对发展创新核系统的各种设想的效率进行比较评估。本文提出了一种基于多准则分析方法的俄罗斯核工业逐步向集中封闭核燃料循环(NFC)双组分核能系统(NES)转移的计算实证方法。同时,鉴于未来的前景不确定,审议了发展国内核工业的各种备选办法。考虑到核能发展的各种趋势,作者确定了以下三组可能的情景。第一组包括“增长”情景,其中机组数量和总装机容量随着时间的推移而增长。第二组假设装机容量增长一段时间后达到平稳水平,此时不存在随时间变化的容量变化。第三组模拟了一个国家核能装机容量在增长之后的下降。为了选择最理想的技术发展方式和评估核能系统的效率,使用了一套有限的选择标准和绩效指标,包括经济、出口潜力、竞争力、SNF和RW的有效管理、天然铀消耗和创新发展潜力。这项工作的一个重要部分是对权重和用于导出标准的输入数据中的不确定性进行详细分析。
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引用次数: 0
On a significant slowing-down of the kinetics of fast transient processes in a fast reactor 快堆中快速瞬态过程动力学的显著减速
Pub Date : 2020-11-20 DOI: 10.3897/nucet.6.60379
G. Kulikov, A. Shmelev, V. Apse, E. Kulikov
The kinetics of nuclear reactors is determined by the average neutron lifetime. When the inserted reactivity is more than the effective delayed neutron fraction, the reactor kinetics becomes very rapid. It is possible to slow down the fast reactor kinetics by increasing the neutron lifetime. The authors consider the possibility of using the lead isotope, 208Pb, as a neutron reflector with specific properties in a lead-cooled fast reactor. To analyze the emerging effects in a reactor of this type, a point kinetics model was selected, which takes into account neutrons returning from the 208Pb reflector to the reactor core. Such specific properties of 208Pb as the high atomic weight and weak neutron absorption allow neutrons from the reactor core to penetrate deeply into the 208Pb reflector, slow down in it, and have a noticeable probability to return to the reactor core and affect the chain fission reaction. The neutrons coming back from the 208Pb reflector have a long ‘deadtime’, i.e., the sum of times when neutrons leave the reactor core, entering the 208Pb reflector, and then diffuse back into the reactor core. During the ‘dead-time’, these neutrons cannot affect the chain fission reaction. In terms of the delay time, the neutrons returning from the deep layers of the 208Pb reflector are close to the delayed neutrons. Moreover, the number of the neutrons coming back from the 208Pb reflector considerably exceeds the number of the delayed neutrons. As a result, the neutron lifetime formed by the prompt neutron lifetime and the ‘dead-time’ of the neutrons from the 208Pb reflector can be substantially increased. This will lead to a longer reactor acceleration period, which will mitigate the effects of prompt supercriticality. Thus, the use of 208Pb as a neutron reflector can significantly improve the fast reactor nuclear safety.
核反应堆的动力学是由中子平均寿命决定的。当插入反应性大于有效延迟中子分数时,反应器动力学变得非常快。通过增加中子寿命来减缓快堆动力学是可能的。作者考虑了在铅冷快堆中使用铅同位素208Pb作为具有特殊性能的中子反射器的可能性。为了分析这种类型反应堆中出现的效应,选择了一个点动力学模型,该模型考虑了从208Pb反射器返回反应堆堆芯的中子。208Pb的高原子量和弱中子吸收等特性使得反应堆堆芯的中子能够深入穿透208Pb反射器,并在反射器中减速,有明显的概率返回反应堆堆芯,影响链式裂变反应。从208Pb反射器返回的中子有很长的“死区时间”,即中子离开反应堆堆芯,进入208Pb反射器,然后扩散回反应堆堆芯的时间总和。在“死区”期间,这些中子不能影响链式裂变反应。在延迟时间方面,从208Pb反射器深层返回的中子接近延迟中子。此外,从208Pb反射器返回的中子数量大大超过了延迟中子的数量。因此,由提示中子寿命形成的中子寿命和来自208Pb反射器的中子的“死区时间”可以大大增加。这将导致一个较长的反应堆加速周期,这将减轻瞬态超临界的影响。因此,使用208Pb作为中子反射体可以显著提高快堆核安全性。
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引用次数: 0
Investigation of the impact of steady-state VVER-1000 (1200) core characteristics on the reactor stability with respect to xenon oscillations 稳态VVER-1000(1200)堆芯特性对氙振荡下反应堆稳定性影响的研究
Pub Date : 2020-11-20 DOI: 10.3897/nucet.6.60464
R. Malkawi, Sergey B. Vygovsky, O. Batayneh
The article presents a method for obtaining an analytical expression for the criterion of stability of a VVER-1000 (1200) reactor with respect to xenon oscillations of the local power in the core, containing an explicit dependence of the criterion ratio coefficients on the arbitrary axial neutron field distribution in steady states of the core. Based on the data of numerical experiments using a full-scale model of the Kalinin NPP power units, the authors present the results of checking the validity of this expression for the reactor stability criterion with respect to xenon oscillations for different NPPs with VVER-1000 (1200) reactors.
本文提出了一种计算VVER-1000(1200)反应堆稳定性判据的解析表达式的方法,其中包含判据比系数与堆芯稳定状态下任意轴向中子场分布的显式依赖关系。基于加里宁核电站全尺寸模型的数值实验数据,作者给出了用VVER-1000(1200)反应堆对不同核电站氙振荡的反应堆稳定性判据的有效性检验结果。
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引用次数: 1
Development of a criterion for assessment of fuel washout during operation of WWER power units WWER发电机组运行过程中燃料损耗评估标准的制定
Pub Date : 2020-11-20 DOI: 10.3897/nucet.6.60559
I. Evdokimov, A. G. Khromov, Petr M. Kalinichev, V. Likhanskii, A. Kovalishin, M. N. Laletin
Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.
核电站在运行过程中可能发生燃料故障。燃料故障可能造成的最严重的后果之一是燃料可能从泄漏的燃料棒中被冲到冷却剂中。在辐照结束后,可靠地检测燃料冲蚀对于处理泄漏燃料组件非常重要。在反应堆运行过程中,通过对冷却剂活性的评估,可以实现对燃料冲刷的检测。为此,134I活动历来用于战时动力装置。然而,观察到的134I活度在运行过程中可能会增加,即使堆芯没有泄漏燃料,并且燃料沉积物是裂变产物释放的唯一来源。本文描述了一个能揭示wwer型反应堆运行过程中由于泄漏燃料棒的燃料冲蚀而导致134I活度升高的判据。讨论了在核电站的一些应用实例。
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引用次数: 0
Repeated measurements and quality of estimates in the analysis of NPP pipeline erosion-corrosion wear 核电厂管道冲蚀磨损分析中的重复测量和评估质量
Pub Date : 2020-11-20 DOI: 10.3897/nucet.6.60459
V. I. Baranenko, O. M. Gulina, S. Mironov, N. L. Salnikov
The article presents a study carried out on carbon steel pipe components subjected to erosion-corrosion wear (ECW). Based on the repeated control data, the authors present the calculated ECW characteristics, i.e., the wall-thinning and rate. It is shown that such estimates contain great uncertainty due to corrosion products deposited on the pipeline inner surface and their migration during operation. In addition, with an increase in the operating time, for example, when the lifetime is extended, the difference between the forecast and the results of control becomes larger. This means that the error in the estimates of the residual lifetime also increases. The study is based on the data of wall thickness measurements of the feedwater pipeline (273×16 mm) and steam pipeline (465×16mm) of nuclear power plants with the VVER-440 reactors, for which a sufficient number of repeated measurements were performed over a large time interval. An analysis is made of the error in estimating the pipeline wall-thinning and ECW rate using Chexal-Horowitz Flow-Accelerated Corrosion (FAC) Model (EKI-02 and EKI-03 software tools). The estimate of the ECW rate according to the above forecast model differs from the estimate according to the current control data by no more than 12.5%, since the corrosion products deposited on the pipeline inner surface wall are leveled at a large time base. When calculating the wall-thinning, due to the obvious filtering of the control data, it is possible to achieve an acceptable accuracy of estimates, i.e., about 16% without upgrading the model.
本文对碳钢管构件进行了冲蚀磨损(ECW)研究。在重复控制数据的基础上,给出了计算得到的ECW特性,即壁厚减薄和速率。结果表明,由于管道内表面沉积的腐蚀产物及其在运行过程中的迁移,这种估算具有很大的不确定性。此外,随着运行时间的增加,例如寿命的延长,预测结果与控制结果之间的差异会变大。这意味着估计剩余寿命的误差也会增加。本研究以VVER-440反应堆核电站给水管道(273×16 mm)和蒸汽管道(465×16mm)的壁厚测量数据为基础,在较长的时间间隔内进行了足够数量的重复测量。分析了利用Chexal-Horowitz流动加速腐蚀(FAC)模型(EKI-02和EKI-03软件工具)估算管道壁薄化和ECW速率的误差。由于沉积在管道内表面壁面的腐蚀产物是在一个大的时间基础上被夷平的,根据上述预测模型估计的ECW速率与根据当前控制数据估计的ECW速率相差不超过12.5%。在计算壁厚减薄时,由于控制数据有明显的过滤作用,在不升级模型的情况下,估计精度可以达到可接受的16%左右。
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引用次数: 1
Precision neutronic calculations of experiments on the neutron transmission through the reflector layers at the BFS critical facilities for expanding the verification database to justify lead cooled fast reactor designs BFS关键设施反射层中子透射实验的精确中子计算,用于扩展验证数据库以证明铅冷快堆设计的合理性
Pub Date : 2020-11-20 DOI: 10.3897/nucet.6.60303
O. Andrianova, Evgeniya S. Teplukhina, Gennady M. Zherdev, Zhanna V. Borovskaya, A. P. Zhirnov
The paper presents the results of the efforts concerned with expanding the verification database and estimating the calculation uncertainty of the power density in the steel reflector of lead cooled fast reactor designs based on experiments performed in different years at the BFS critical assemblies by analyzing and revising earlier calculation and experimental studies on the transmission of neutrons through the steel reflector layers. The discussion includes experiments at the BFS-66 critical assembly to model neutron and photon fluxes in the reactor core shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 critical assemblies to model the transmission of neutrons and gamma quanta through the reflector layers of various materials. The information provided in earlier materials with the descriptions of the above experiments has been analyzed and expanded through respective data required to prepare precision calculation models for Monte-Carlo neutronic codes. Precision neutronic models have been developed based on actualized and updated data with a detailed description of the BFS heterogeneous structure and experimental devices, and test calculations have been carried out to confirm their efficiency. The calculations of key neutronic characteristics measured at the BFS-66, -64 and -80-2 assemblies were performed using codes based on the Monte Carlo method (MCU-BR, MCNP, MMK-RF, MMK-ROKOKO) with BNAB-RF and MDBBR50 neutron data and the ROSFOND evaluated neutron data library. The developed precision calculation neutronic models of the experiments discussed can be used to justify lead cooled fast reactor designs, to verify neutronic codes and neutron data, and to evaluate the associated uncertainties.
本文介绍了在不同年份对铅冷快堆临界组件进行实验的基础上,通过分析和修正早期关于中子穿过钢反射层的计算和实验研究,扩充验证数据库和估算铅冷快堆钢反射层功率密度计算不确定性的工作成果。讨论包括在BFS-66临界组件上模拟反应堆堆芯屏蔽成分中的中子和光子通量的实验,以及在BFS-64和BFS-80-2临界组件上模拟中子和伽马量子通过各种材料反射层的传输的实验。对前面材料中所提供的信息以及上述实验的描述进行了分析,并通过相应的数据进行了扩展,以制备蒙特卡罗中子码的精确计算模型。根据实际和更新的数据,详细描述了BFS的异质结构和实验装置,建立了精确的中子模型,并进行了试验计算,以验证其有效性。采用基于蒙特卡罗方法(MCU-BR、MCNP、MMK-RF、MMK-ROKOKO)的代码,利用BNAB-RF和MDBBR50中子数据和ROSFOND评估中子数据库,对BFS-66、-64和-80-2组件测量的关键中子特性进行了计算。所建立的精确计算中子模型可用于验证铅冷快堆的设计,验证中子代码和中子数据,并评估相关的不确定性。
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引用次数: 0
16Cr-19Ni steel swelling at dose rates from 1×10-8 to 1.6×10-6 dpa/s 16Cr-19Ni钢在1×10-8 ~ 1.6×10-6 dpa/s剂量率范围内膨胀
Pub Date : 2020-11-18 DOI: 10.3897/nucet.6.60371
E. Kinev
The article presents the first data on EK-164ID steel swelling after operational irradiation in a fast nuclear reactor in the temperature range of 370–630 °C and maximum damaging doses of 66–77 dpa. The dose accumulation rate along the cladding tubes made of this material was 1×10-8–1.6×10-6 dpa/s. The swelling was determined by the hydrostatic weighing method with an error of no more than 0.5%. The results obtained were analyzed depending on the irradiation parameters and in comparison with the 16Cr-15Ni grade material. The objectives of the study were to estimate the characteristic values of the maximum swelling temperature and dose as well as to calculate the average material swelling rate at the working temperature of irradiation, the incubation period for the onset of swelling, and the stationary swelling rate. It was found that the tube samples, characterized with austenite grain sizes of 9–12 μm before irradiation, have an average swelling rate of 0.035–0.05 %/dpa after reaching the maximum damaging doses of 66–77 dpa (at a rate of (1–1.5)×10-6 dpa/s) and not more than 0.035%/dpa at doses less than 20 dpa (at a rate of 5×10-7 dpa/s). The characteristic maximum swelling temperature of the studied material is in the range of 430–500 °C. The characteristic maximum swelling dose is in the range of 61–72.5 dpa or 70–80% of the maximum accumulated dose. The incubation stationary swelling period for the material is 30 dpa. The stationary swelling rate is 0.1% /dpa. The radiation resistance characteristics of the studied material have an advantage over those for 16Cr-15Ni grade cladding materials under similar irradiation conditions and a similar structural state, which inherits grain sizes of 9–14 μm during the tube processing.
本文首次报道了EK-164ID钢在快中子反应堆中370 ~ 630℃、66 ~ 77 dpa运行辐照后的膨胀现象。沿该材料制成的包层管的剂量累积速率为1×10-8-1.6×10-6 dpa/s。溶胀量采用静压称重法测定,误差不大于0.5%。根据辐照参数对所得结果进行了分析,并与16Cr-15Ni级材料进行了比较。本研究的目的是估计最大溶胀温度和剂量的特征值,并计算辐照工作温度下物质的平均溶胀率、溶胀开始的潜伏期和固定溶胀率。结果表明,辐照前奥氏体晶粒尺寸为9 ~ 12 μm的管状试样,在达到最大损伤剂量66 ~ 77 dpa(速率为(1 ~ 1.5)×10-6 dpa/s)时,平均溶胀率为0.035 ~ 0.05% /dpa,在剂量小于20 dpa(速率为5×10-7 dpa/s)时,溶胀率不大于0.035%/dpa。所研究材料的最大膨胀温度为430-500℃。典型的最大肿胀剂量在61-72.5 dpa或最大累积剂量的70-80%之间。材料的孵育稳定膨胀期为30dpa。固定膨胀率为0.1% /dpa。该材料的耐辐射性能优于相同辐照条件下的16Cr-15Ni级包层材料,其在管加工过程中继承了9 ~ 14 μm的晶粒尺寸。
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引用次数: 0
Basic models and approximation for the engineering description of the kinetics of the oxide layer of steel in a flow of heavy liquid metal coolant under various oxygen conditions 在各种氧条件下,重液态金属冷却剂流动中钢氧化层动力学工程描述的基本模型和近似
Pub Date : 2020-11-16 DOI: 10.3897/nucet.6.59068
A. Avdeenkov, O. I. Achakovsky, V. Ketlerov, V. Kumaev, A. I. Orlov
The article presents the results of corrosion processes, kinetics and changes in the oxide layer modeling using MASKA-LM software complex. The complex is intended for a numerical simulation of three-dimensional non-stationary processes of mass transfer and interaction of impurity components in a heavy liquid metal coolant (HLMC: lead, lead-bismuth). The software complex is based on the numerical solution of coupled three-dimensional equations of hydrodynamics, heat transfer, formation and convective-diffusive transport of chemically interacting components of impurities. Examples of calculations of mass transfer processes and interaction of impurity components in HLMC, formation of protective oxide films on the surfaces of steels are given to justify the coolant technology.
本文介绍了使用MASKA-LM软件对腐蚀过程、动力学和氧化层变化进行建模的结果。该配合物用于重液态金属冷却剂(HLMC:铅,铅-铋)中杂质组分的三维非平稳传质和相互作用过程的数值模拟。该软件是基于流体力学、传热、形成和杂质化学相互作用组分对流扩散输运的耦合三维方程的数值解。文中给出了计算HLMC中杂质组分的传质过程和相互作用、钢表面氧化保护膜形成的实例,以证明冷却剂技术的合理性。
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引用次数: 2
Codes of new generation for safety justification of power units with a closed nuclear fuel cycle developed for the “PRORYV” project 为“PRORYV”项目开发的具有封闭核燃料循环的新一代动力装置安全论证规范
Pub Date : 2020-11-06 DOI: 10.3897/nucet.6.54710
L. Bolshov, V. Strizhov, N. Mosunova
The article describes the status of development of codes of new generation for the “PRORYV” Project by the end of 2019: twenty-five commercial-grade software products to justify design solutions and safety of power units with fast neutron reactors and liquid metal coolant (sodium and lead) in a closed nuclear fuel cycle. The developed system of codes is multi-physical and multi-scale that allows performing both calculations of the whole installations and high precision calculations of their individual elements. The developed codes offer unique features. Twelve developed codes have already been certified by Rostechnadzor, and six more have been submitted for certification. In addition to creating the software products, a large-scale work is being carried out to conduct experimental studies for code validation that meet modern requirements imposed by the codes: unique measurement techniques have been created; experimental data on flow characteristics of heavy liquid metal coolant (HLMC) in a fuel assembly simulator have been obtained, as well as of “gas-HLMC” interphase interaction after inert gas injection in HLMC and characteristics of heat exchange between the inert gas and HLMC. The results are already used for validation of system and CFD codes used in the “PRORYV” Project.
文章描述了到2019年底“PRORYV”项目新一代代码的开发现状:25个商业级软件产品,用于证明快中子反应堆和封闭核燃料循环中液态金属冷却剂(钠和铅)的动力装置的设计解决方案和安全性。开发的代码系统是多物理和多尺度的,允许对整个装置进行计算,也允许对单个元件进行高精度计算。开发的代码提供了独特的功能。Rostechnadzor已经认证了12个开发的代码,另有6个已提交认证。除了创建软件产品外,正在进行一项大规模的工作,以进行符合代码所施加的现代要求的代码验证实验研究:创建了独特的测量技术;获得了重液态金属冷却剂(HLMC)在燃料组件模拟器中的流动特性实验数据,以及在HLMC中注入惰性气体后的“气-HLMC”界面相互作用和惰性气体与HLMC之间的热交换特性实验数据。研究结果已经用于验证“PRORYV”项目中使用的系统和CFD代码。
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引用次数: 4
期刊
Nuclear Energy and Technology
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