Global Nuclear Safety最新文献_第2页

Potential and possibility of extending the campaign to increase VVER-1000 power units on sliding distribution of the secondary circuit 扩大在二次回路滑动分配上增加VVER-1000功率单元的运动的潜力和可能性

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-08

A. S. Grachev, A. Lapkis, A. Smolin

{"title":"Potential and possibility of extending the campaign to increase VVER-1000 power units on sliding distribution of the secondary circuit","authors":"A. S. Grachev, A. Lapkis, A. Smolin","doi":"10.26583/gns-2023-02-08","DOIUrl":"https://doi.org/10.26583/gns-2023-02-08","url":null,"abstract":"The paper considers the efficiency and possibility of implementing into the operational practice of VVER-1000 power units the campaign extension mode at the sliding pressure of the second circuit. The history of research of Russian nuclear scientists in terms of extending the campaign of VVER reactors is considered. It is shown that reducing the pressure of fresh steam within the permissible limits will release additional reactivity due to the release of the temperature effect and generate additional energy during the operation of the unit on the power effect. A mathematical model of changes in the main parameters of the reactor plant has been compiled, including the reactivity balance equations, the Stodola-Flugel equation for the turbine, the equations of thermal balance and heat transfer in the steam generator. The calculation of the expected increase in electricity generation for modern fuel loads of VVER-1000 reactors has been performed. The degree of compliance of the proposed regime with the current technological limitations of the VVER-1000 «large series» power unit is analyzed. The results of the analysis were confirmed by numerical experiment on a multifunctional mode analyzer of the VVER-1000 reactor unit developed by the MEPhI Research Institute, built on the basis of the ENIKAD medium and the certified diffusion neutron-physical code PROSTOR. Conclusions are drawn about the possibility of introducing the proposed regime into the operational practice of modern VVER-1000 power units.","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121504434","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Input data to match model and prototype states of WWER NPP 输入数据以匹配WWER核电厂的模型和原型状态

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-07

I. A. Molev, D. A. Solovyev, A. L. Lobarev, D. Plotnikov, H. A. Tahash, E. V. Chernov

{"title":"Input data to match model and prototype states of WWER NPP","authors":"I. A. Molev, D. A. Solovyev, A. L. Lobarev, D. Plotnikov, H. A. Tahash, E. V. Chernov","doi":"10.26583/gns-2023-02-07","DOIUrl":"https://doi.org/10.26583/gns-2023-02-07","url":null,"abstract":"The paper considers some possible input data of NPP power unit model for solving the problem of model state matching with prototype state. In the course of work on matching the state of NPP model with the prototype, a numerical experiment using full-scale modeling complex PROSTOR was conducted. Its essence was to obtain steady states of the model with capacity from 94% to 104% with a step of 0,05%. For modeling we used data of Kalinin NPP unit 3 loading 4 on 248 effective days. When analyzing the results of the numerical experiment, two effects were found. One of them, expressed in the bend of the derivative of the obtained graph, is connected with reaching the upper limit of the possibility to regulate the pressure in the main steam collector. The regulating valves on the turbine generator are the executing mechanism of pressure regulator in main steam collector. At the break of the graph they open at 100%, which causes further inability to regulate pressure. The pressure in the main steam collector starts to increase. This effect corresponds to the real processes occurring at the NPP. In this case, as one of the model tuning parameters, it is proposed to use the adaptive coefficient to the concentration of boric acid in the reactor, for the control of which, it is proposed to use a mathematical boric regulator. It was also shown that it is impossible to accurately control the pressure in the reactor due to the peculiarities of operation of the volume compensator electric heaters automatics. In this regard, it was proposed to use an additional mathematical volume compensator electric heaters regulator to match the state of the model with the prototype. These regulators have been successfully tested as part of the PROSTOR software package to match the states of the model and the prototype.","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"74 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129701461","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Arc stability study at the limit mode of welding with coated electrodes 涂敷电极焊接极限模式电弧稳定性研究

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-04

S. M. Burdakov, O. Tsuverkalova, Yu. V. Zayarov, L. V. Postoy, V. G. Tkachev

{"title":"Arc stability study at the limit mode of welding with coated electrodes","authors":"S. M. Burdakov, O. Tsuverkalova, Yu. V. Zayarov, L. V. Postoy, V. G. Tkachev","doi":"10.26583/gns-2023-02-04","DOIUrl":"https://doi.org/10.26583/gns-2023-02-04","url":null,"abstract":"This paper shows the results of experiments to study the effect of the auxiliary voltage with frequency fi= 40000Hz on the stability of arc discharge at the limit mode of the welding process with single, coated electrodes. This effective method of connecting rolled and prefabricated elements of metal structures has found widespread use in their installation, manufacture and repair in the construction of various facilities: industrial and civil purposes, thermal and nuclear power facilities for responsible purposes. In this regard, the increase in the quality and technological, strength parameters of welds, the productivity of the manufacturing metal structure process, is largely associated with welding modes and an increase in the stability of arc discharge as an important element of the system: \"power source-arc-welding tub-product.\" Improving the performance of the welding process under consideration is particularly important under mounting conditions in conjunction with its stability is an important task on the agenda. When performing the study, the following criteria and their values regarding the stability of the welding arc were taken into account, such as: lexpl - rupture length in mm, coefficients of variation: KV(Iwd) - welding current, KV(Uwd) - voltage, qt - constant time, as well as their numerical values were obtained. The obtained numerical values of the above criteria were obtained based on the analysis of graphs (oscillograms) of the dependence of the electric current (Iwd, A) and the voltage (Uwd, V) of the arc on the time of the process duration (t, sec). The obtained results of the analysis indicate an increase in stability, a decrease in inertia of the deviation of the resistance (conductivity) of the welding arc when the auxiliary voltage is applied with a frequency of fi = 40000Hz at the limit mode of welding with coated electrodes.","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"49 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134392472","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Diagnostics of the drive of the control and protection system of the reactor plant ark type 反应器厂柜式控制与保护系统驱动的诊断

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-09

E. Abidova, P. V. Povarov, V. Popov, O. Pugacheva, V. Shpicer

{"title":"Diagnostics of the drive of the control and protection system of the reactor plant ark type","authors":"E. Abidova, P. V. Povarov, V. Popov, O. Pugacheva, V. Shpicer","doi":"10.26583/gns-2023-02-09","DOIUrl":"https://doi.org/10.26583/gns-2023-02-09","url":null,"abstract":"The drive of the reactor control and protection system (CPS) is a system of normal operation, important for safety. Malfunctions of the drives of the ARC type CPS are often the initial events for accidents leading to unauthorized downtime. The currently existing standard methods for monitoring the parameters of the CPS do not allow for a reliable assessment of the condition of the drives. In this paper, it is proposed to implement an approach that provides an increase in sensitivity when recognizing the states of the drives of the ARC control system by processing the initial signals, which are vibroacoustic signals registered on the lid of the upper unit of the VVER reactor. A procedure for processing diagnostic signals using singular spectrum analysis is proposed. The proposed approach differs from the known ones by the presence of a scaling procedure, which is implemented by multiplying the eigenvalue matrices by the gankelized matrices of the source data. Due to scaling, the necessary sensitivity increase is provided. The expected increase in sensitivity is based on the fact that the eigenvalues reflect the structure of the signals, which changes significantly under the influence of the defect. The proposed method was used to process vibroacoustic signals registered on the lid of the upper block of the VVER reactor unit. Experimental studies were carried out at the test stand of Izhorskiye Zavody OJSC. The results of processing experimental data indicate a high quality of diagnosis. The hypothesis was clearly confirmed that the difference in the structure of diagnostic signals of serviceable and faulty equipment can manifest itself in the eigenvalues of the gankel signal matrices. The approach proposed in the article to the processing of diagnostic signals is easily amenable to automation and can be implemented in the development of a diagnostic system for the drive of an ARC-type CPS.","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"19 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114132642","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Features of applying safeguards to Optimized Floating Power Unit spent nuclear fuel handling 优化浮动动力机组乏核燃料处理保障措施的应用特点

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-01

S. Dushev, I. V. Shmelev, P. Gorbunov, N. Salnikova, A. D. Abrosimov, V. Malev, T. A. Tagirova

{"title":"Features of applying safeguards to Optimized Floating Power Unit spent nuclear fuel handling","authors":"S. Dushev, I. V. Shmelev, P. Gorbunov, N. Salnikova, A. D. Abrosimov, V. Malev, T. A. Tagirova","doi":"10.26583/gns-2023-02-01","DOIUrl":"https://doi.org/10.26583/gns-2023-02-01","url":null,"abstract":"After the Floating Power Unit Akademik Lomonosov with the KLT-40S reactor plant had been successfully commissioned, the next step for ROSATOM was to develop optimized floating power units (OFPU). OFPUs powered by the RITM reactor plant are an export-oriented product, and they designed to operate on sites outside Russia. OFPU engineering solutions are based upon the practical design and operation experience in nuclear-powered icebreakers and upon problem solving to assure safety for operation of nuclear plants under severe northern conditions. At the nuclear fuel enrichment lower than 20%, the refueling interval for the OFPU may be up to 10 years depending on Loadings. The RITM reactor refueling process has a number of differences compared to the refueling process of other nuclear power plants fitted with water-cooled water-moderated power reactors. RITM plants have never been placed under the IAEA Safeguards, and capabilities and features of applying the IAEA Safeguards to such plants require an additional analysis. The developed operation model assumes that the OFPU be operated on a deployment site for a long period, and all fuel handling operations will be done at a specialized facility in the Russian Federation (RF). This paper is discussing the major fuel handling features in the life cycle of the OFPU fitted with RITM-200M—these are the critical features in terms of the IAEA Safeguards. Also, the paper is providing the designers’ perspective on solving the key non-proliferation problems through adapting the Safeguards approaches with account of the OFPU technical features.","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"41 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123460814","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Improvement of methods of algolization and biomelioration of the Rostov NPP cooling pond and the near dam part of the Tsimlyansk reservoir 罗斯托夫核电站冷却池和茨姆利扬斯克水库近坝部分藻化和生物净化方法的改进

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-02

O. Y. Gorskaya

引用次数: 0

Predictive diagnostics of RBMK reactor flowmeters RBMK反应器流量计的预测诊断

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-03

A. Zagrebaev, E. O. Popov

{"title":"Predictive diagnostics of RBMK reactor flowmeters","authors":"A. Zagrebaev, E. O. Popov","doi":"10.26583/gns-2023-02-03","DOIUrl":"https://doi.org/10.26583/gns-2023-02-03","url":null,"abstract":"At the moment, there are 8 rectors of the RBMK type in operation. The safety of their operation is ensured by monitoring devices for key indicators. In this article, the main attention is paid to the control of the coolant flow through the fuel channel, carried out with the help of ball flow meters of the STORM-32M and STORM-8A types. During operation, there is a gradual wear of the raceway and the ball, which leads to the appearance of a negative error in the device, in other words, the readings become underestimated. In this regard, as well as due to the widespread use of ball flowmeters at nuclear power plants, the task of diagnosing the state of flowmeters is urgent. One of the approaches to determining the flow rate of the coolant is an approach based on the measurement and use of information about the activity of the coolant by the tightness control system of the fuel element shells (SKGO). The use of this radiation approach to control the flow of coolant has its advantages and difficulties. The advantage is that real experimental information about the presence of a coolant in the fuel channel is involved, and the difficulties are associated with receiving and processing the signals of the SCGO in real time and, consequently, with low efficiency of solving the problem. In this paper, we propose a method for predictive diagnostics of the flow meter condition using the design features of the multiple forced circulation circuit (CMPC) of the reactor. The adaptive parameter of the model was selected as a diagnostic parameter. Studies have shown that the increase in time variance in experimental flow values has a clear effect on the selected diagnostic feature , as a result of which it can be said that the observation of this kind of variance change in experimental data can serve as a sign of a malfunction of the flow meter.","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"67 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133634464","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Prediction of test results to determine the hydraulic resistance coefficient of an axially symmetrical check valve 预测试验结果，以确定轴对称止回阀的水力阻力系数

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-06

S. Kuzin, S. B. Kravets

{"title":"Prediction of test results to determine the hydraulic resistance coefficient of an axially symmetrical check valve","authors":"S. Kuzin, S. B. Kravets","doi":"10.26583/gns-2023-02-06","DOIUrl":"https://doi.org/10.26583/gns-2023-02-06","url":null,"abstract":"The urgency of the paper is determined by the necessity to predict the results of tests for determining the coefficient of hydraulic resistance of axially symmetrical check valve DN 100-600 with operating pressure up to 20 MPa and operating temperature from plus 50C to plus 3500C according to TS 6981-254-08847871-07. This paper suggests to reduce the volume of tests by excluding the tests for intermediate samples of valves (DN 100, DN125, DN 200, DN 300, DN 500, DN 600) situated in DN between the valve type models (DN 100, DN 175, DN 400) and replaced the tests by calculations in the programme module ANSYS CFX of the calculation complex ANSYS. The results of the tests for determining the coefficient of hydraulic resistance are given. An estimate of hydraulic resistance coefficient values is offered and recommendations for its determination are given. The experimental determination of hydraulic characteristics was carried out in the region of quadratic resistance and in absence of cavitation (the tests were done on water). The region of quadratic resistance comes at Reynolds numbers Re ≥ 2 × 104 for check valve of nominal diameters DN ≤ 250. The region of quadratic resistance ReKB is determined in the process of experimental determination of drag and flow capacity factors for valves of nominal diameters DN ≥ 250. The procedure of determination of ReKB is described in 7.2.3 of GOST 55508-2013. These test results for determination of hydraulic resistance coefficient for valves DN 150, DN 175, DN 400 according to TU 6981-254-08847871-07 can be used when developing acceptance test programs to set the number of tested samples according to the design features and number of versions of check valves (DN 100, DN125, DN 200, DN 300, DN 500, DN 600), replacing the testing by calculation.","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"28 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126669617","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Presentation of electrical control results by electrophysical chromatography method 电物理色谱法电气控制结果的介绍

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-05

V. Surin, A. S. Shcherban, A. Shcherbakov, M. B. Ivanyi, M. Zhidkov, S. Tomilin, A. Kozlov

{"title":"Presentation of electrical control results by electrophysical chromatography method","authors":"V. Surin, A. S. Shcherban, A. Shcherbakov, M. B. Ivanyi, M. Zhidkov, S. Tomilin, A. Kozlov","doi":"10.26583/gns-2023-02-05","DOIUrl":"https://doi.org/10.26583/gns-2023-02-05","url":null,"abstract":"A method of electrophysical chromatography was proposed to visualize the results of electrical non-destructive testing for the first time. Surface potentiograms usually used to represent the results of scanning contact potentiometry are replaced by volumetric images of structural inhomogeneities. The alternative has obvious advantages as useful additional information can drastically change the results of testing which is demonstrated by the example of a weld overlay. Methodological recommendations for constructing volumetric images of defects are formulated and implemented. The causes of self-shielding of structural inhomogeneities at different levels of fixation which are due to the mutual spatial overlap of microregions emitting waves of elastic stresses are discussed. The biggest problems in identifying defects for NDT are their complete or partial screening. The electrophysical chromatography method requires at least one experimental array obtained from the following measurements: double scanning of the surfaces of the test object; double simultaneous scanning of the surfaces of the test object; double simultaneous scanning of the surfaces of the test object using a radiation synchronizer in time or frequency. In the case of scanning along one (external or internal) surface, the coordinates of defects are determined from the corresponding cross sections and the calculated values of their depth using frequency and time-frequency analysis","PeriodicalId":113403,"journal":{"name":"Global Nuclear Safety","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130801333","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

A model of assessing the economic efficiency of introducing Rosatom Production System projects to upgrade the NPP equipment 引进俄罗斯国家原子能公司生产系统项目升级核电站设备的经济效益评估模型

Global Nuclear Safety Pub Date : 2023-06-21 DOI: 10.26583/gns-2023-02-11

I. A. Ukhalina, M. Golovko, S. P. Agapova, A. Antsibor, N. A. Efimenko, Z. S. Rogacheva

引用次数: 0