Nuclear Engineering and Design最新文献

{"title":"Conditions for economic efficiency of latent heat thermal energy storage systems at nuclear power plants","authors":"Valery Yurin ,&nbsp;Michael Garievsky ,&nbsp;Daniil Anoshin","doi":"10.1016/j.nucengdes.2024.113581","DOIUrl":"10.1016/j.nucengdes.2024.113581","url":null,"abstract":"<div><p>In conditions of uneven schedules of electricity consumption and construction of power plants using renewable energy sources, nuclear power plants will participate in regulating the unevenness of daily energy consumption. At the same time, nuclear power plants have high economic feasibility of operating with a maximum installed capacity utilization factor due to significant capital investments at a relatively low fuel price. The use of nuclear power plants to cover uneven energy consumption will reduce the possibility of a return on investment, which will only be realized with greatly inflated capacity charges. At the same time, the annual operating costs will increase significantly. A possible decrease in the economic efficiency of a nuclear power plant is due to the fact that, at the request of the system operator of the energy system, the NPP will be to reduce electricity generation during off-peak hours or sell electricity on the market at significantly lower prices. In this regard, the problem of increasing economic efficiency of nuclear power plants under conditions of uneven daily energy consumption becomes particularly relevant. The solution to the problem could be the accumulation of excess energy and its subsequent use. Previously, the authors developed a method of additional redundancy of auxiliary needs of nuclear power plants based on the use of low-power steam turbines. In the present paper, schemes for increasing efficiency of using low-power steam turbines at nuclear power plants when regulating the load unevenness in the power system using thermal energy storage systems based on the latent heat thermal energy storage are developed. The design characteristics of the accumulator as part of a nuclear power plant are determined to ensure operation of the heat storage system throughout the entire period of energy consumption. In this research, the lithium nitrate was considered as a phase change material. The technical and system conditions under which the maximum economic effect is achieved have been determined. The boundary conditions under which a return on investment is achieved are shown.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142167414","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Prediction of and genetic algorithm optimization on data induced uncertainty reduction with the use of an integral experiment","authors":"Alexander Amedeo DePillis,&nbsp;Sandra Bogetic,&nbsp;Vladimir Sobes","doi":"10.1016/j.nucengdes.2024.113570","DOIUrl":"10.1016/j.nucengdes.2024.113570","url":null,"abstract":"<div><p>This paper verifies a method for predicting the uncertainty reduction on a parameter of interest in a target application system from adjusting nuclear data based on an integral experiment and developed a procedure for optimizing integral experiments for particular applications. This paper also discusses the expansion of a global optimization software package, Gnowee, to optimize the design of the Flexible Neutron Source (FNS) at the University of Tennessee Knoxville, in combination with MCNP, for the purpose of reducing uncertainties from nuclear data on applications’ parameters of interest. The optimization process considered various isotopes and reactions, and the final design reduced the number of sodium plates almost to the minimum allowed while using a combination of fuel plates and empty spaces. The final design is predicted to be able to reduce the cross section induced uncertainty on a generic sodium fast reactor by 15–20 % on the k-eigenvalue.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142163170","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"An efficient method for input uncertainty propagation in CFD and the application to buoyancy-driven flows","authors":"Ruiyun Ji ,&nbsp;Stephan Kelm ,&nbsp;Markus Klein","doi":"10.1016/j.nucengdes.2024.113560","DOIUrl":"10.1016/j.nucengdes.2024.113560","url":null,"abstract":"<div><p>Severe accident scenarios address the release of large amounts of hydrogen and steam to the containment. The formation of a flammable gas cloud could lead to a combustion and even failure of containment structures. In order to support the hydrogen mitigation method development, a detailed understanding of the gas transport and mixing process is crucial. Efforts in terms of numerical simulations such as Computational Fluid Dynamics (CFD) models have been made, which allow to investigate the complex 3D gas mixing process. One of the uncertainty sources that challenge the reliability of CFD validation results is the input uncertainty. It was assessed efficiently using the deterministic sampling method, which requires e.g., in the present case only eight binary samples for seven uncertain input parameters. However, the lean number of samples makes the direct derivation of a probability density function as well as a 95% confidence interval impossible. The assumption of a normal distribution does not always yield convincing and physically consistent output uncertainty bands, in particular for measurements inherent to oscillations. In this context, a new method has been proposed, which enables the generation of reasonable pseudo-samples without additional CFD simulations and the derivation of 95% confidence interval through the statistical analysis on these pseudo-samples. It was assessed against the Monte Carlo sampling method with a simple test case and confirmed an improved prediction. This method has been applied to the large scale application-oriented validation case THAI-TH32 in this work, in order to assess the impact of input uncertainties on the CFD results.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0029549324006605/pdfft?md5=bd613eb8d09326f56106a6885c26d27b&pid=1-s2.0-S0029549324006605-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142163172","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"State feedback control of HPR1000 average coolant temperature based on dominant pole","authors":"Ziqi Fan,&nbsp;Xianshan Zhang,&nbsp;Kaiyang Zheng,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.nucengdes.2024.113569","DOIUrl":"10.1016/j.nucengdes.2024.113569","url":null,"abstract":"<div><p>Control of nuclear power plant is still based on the traditional PID control system, which is difficult to obtain high control quality in the process of a wide range of load changes. To effectively use the measurable information of the system and consider the constraints, state feedback control based on the dominant pole method is proposed for the average coolant temperature control of HPR1000. The control system is divided into two parts: one part is a feedback branch, which realizes the state feedback by using the measurable system state quantity including the core inlet temperature, the core outlet temperature and the reactor power, and at the same time introduces the integral link to reduce the steady-state error; the other part is a feedforward branch, which uses the nominal load change to make feedforward compensation to improve the control performance of load tracking. At the same time, Particle Swarm Optimization (PSO) method is used to optimize the controller parameters, and the dominant pole meeting the requirements is obtained. The control performance under different working conditions is verified on the HPR1000 model. The test results show that the state feedback control can effectively improve the setpoint tracking ability and anti-disturbance ability.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0029549324006691/pdfft?md5=ea054a99fc75dcd24f5692a939d1f2dd&pid=1-s2.0-S0029549324006691-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142163171","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study on eutectic reaction between lower head of reactor pressure vessel and molten pool using simulant materials","authors":"Zhihong Xiong ,&nbsp;Weiqi Cheng ,&nbsp;Kaijun Ou ,&nbsp;Songbai Cheng ,&nbsp;Bing Tan","doi":"10.1016/j.nucengdes.2024.113584","DOIUrl":"10.1016/j.nucengdes.2024.113584","url":null,"abstract":"<div><p>The eutectic reaction between core materials is an important behavior during reactor severe accidents because it will affect the accident progression. To comprehend the eutectic reaction between molten corium and the lower head of the reactor pressure vessel, this study conducted a series of simulated experiments. Liquid bismuth (Bi) and lead (Pb) rods, chosen for their relatively low eutectic point of 398.5 K, were utilized in a self-designed apparatus. For a more comprehensive understanding, various experimental parameters, including the temperature of Bi pool (548–578 K), temperature of Pb rod (548–578 K), charged gas pressure (0.002–0.006 MPa), weight of load block (3.3–7.3 kg) and style of baffles (type A, type B) were varied. Upon analyzing the experimental observations and quantitative data gathered, we investigated the influence of experimental conditions on the reaction rate. It is found that the temperature of the Bi pool exerts a notably positive influence on the eutectic reaction rate constant, regardless of the baffle pattern and the initial Pb rod temperature, while the role of Pb rod temperature seems to be insignificant. In experiments employing baffle A, both the weight of the load block and gas pressure significantly impact the reaction rate constant. Conversely, in the case of baffle B, their impact is insignificant. Knowledge and evidence gained from this study will be applied to future preparation of high-temperature experiments that involve actual reactor materials. Furthermore, this study furnishes experimental data aimed at enhancing and validating the eutectic reaction related model within the severe accident analysis code for water-cooled reactors.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142163173","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Helical coil steam generator stability experiments with the MOTEL test facility","authors":"Vesa Riikonen ,&nbsp;Joonas Telkkä ,&nbsp;Virpi Kouhia ,&nbsp;Markku Puustinen ,&nbsp;Giteshkumar Patel ,&nbsp;Lauri Pyy ,&nbsp;Antti Räsänen ,&nbsp;Eetu Kotro ,&nbsp;Juhani Hyvärinen ,&nbsp;Afeef Murad","doi":"10.1016/j.nucengdes.2024.113566","DOIUrl":"10.1016/j.nucengdes.2024.113566","url":null,"abstract":"<div><p>Small modular reactors (SMRs) are under extensive development globally. Some SMR concepts have design features that are rare in traditional pressurized water reactors (PWRs). One such feature is a helical coil steam generator which differs from traditional horizontal and vertical inverted U-tube steam generators in several ways. The helical coil steam generator is a once-through design where the primary side flow runs in the shell side and the secondary side flow runs inside the tubes and can generate superheated steam. Boiling instabilities in helically coiled tubes are a crucial research question due to their potentially negative impact on steady plant operation. The MOTEL (MOdular TEst Loop) facility at LUT University is a model of an integral SMR with a helical coil steam generator representing an integral pressurized water reactor. A large variation of core power and feedwater flow values were tested to map MOTEL operating conditions in which the steam production in the helical coil steam generator is stable.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0029549324006666/pdfft?md5=78133a222df5150e8bba2360cfb6b2c4&pid=1-s2.0-S0029549324006666-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142151756","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Operation experience of a high-temperature fluoride salt test facility (FLUSTFA): Issues identified and paths forward","authors":"Sheng Zhang,&nbsp;Shuai Che,&nbsp;Adam Burak,&nbsp;Xiaodong Sun","doi":"10.1016/j.nucengdes.2024.113568","DOIUrl":"10.1016/j.nucengdes.2024.113568","url":null,"abstract":"<div><p>Fluoride-salt-cooled High-temperature Reactors (FHRs) are promising Generation IV nuclear reactors, with a passive decay heat removal system serving as one of their key design features. A high-temperature FLUoride Salt Test FAcility (FLUSTFA) was designed and constructed to perform both integral-effect tests, such as validating the design of a scaled-down passive decay heat removal system, and separate-effect tests, including evaluating the thermal–hydraulic performance of compact heat exchangers. FLUSTFA utilizes FLiNaK (LiF-NaF-KF: 46.5–11.5–42 mol%) as the working fluid and operates up to 700 °C near the atmospheric pressure. It is comprised of a reservoir tank for melting and storing the salt, a primary molten salt loop that simulates the reactor primary coolant system, a secondary molten salt loop that represents the passive decay heat removal system, an air loop, and a chilled water loop, all of which are thermally coupled via heat exchangers. Several shakedown tests were carried out using high-temperature nitrogen and FLiNaK salt as the working fluid. A few issues were identified during initial operation of FLUSTFA at 550–600 °C, such as hydrogen fluoride generation and leakage, localized hot spots where excessive heat loss occurs, molten salt pump malfunctions, abnormal readings from ultrasonic flow meters, and blockage of salt charging and recycling lines. In addressing all these issues, paths forward have been successfully identified and implemented. The lessons learned are valuable in improving future design and construction of high-temperature molten salt systems for molten salt reactors, fusion reactors, and next-generation concentrated solar power plants.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142151656","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"The safety criteria of high flux research reactor with downward parallel channels: onset of flow instability and critical heat flux","authors":"Zhuang Wang,&nbsp;Gan Zhu,&nbsp;Heng Xie","doi":"10.1016/j.nucengdes.2024.113572","DOIUrl":"10.1016/j.nucengdes.2024.113572","url":null,"abstract":"<div><p>This paper focuses on safety criteria for high flux reactors with parallel downflow channels: critical heat flux (CHF) and onset of flow instability (OFI). Sudo93 CHF correlations, a set of correlations widely used in high flux research reactors, are used to predict CHF. Flow excursion is especially important for high flux reactors as it can lead to premature CHF. Therefore, OFI prediction is a must for high flux reactors. In this paper, the database of Sudo93 CHF correlations is studied and analyzed. It’s found that the data in the CHF database have the possibility of being affected by flow excursion. In addition, Sudo93 CHF correlations are used to evaluate the collected OFI experimental data. It is shown that Sudo93 CHF correlations can predict downflow OFI under certain conditions. The reason is given by comparing Sudo93 CHF correlations with OFI prediction correlations. The –33% uncertainty of Sudo93 CHF correlations can also be accounted for from the perspective of OFI. Furthermore, this paper analyzes the deviations between the predicted and measured values in Sudo CHF database. It is confirmed that OFI influences the lower limit of Sudo93 CHF correlations. Eventually, this paper proposes the improved downflow Sudo93 CHF correlations by introducing the flow instability factor, which improves the economic benefits while ensuring safety. When the improved Sudo93 CHF correlations are used for high flux reactors, additional OFI prediction is unnecessary. This further reduces unnecessary safety margin. Therefore, this study has great significance for the design and safety analysis of high flux research reactors.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142151755","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"TRACE code core reflood thermal-hydraulics phenomena benchmarking against the NRC–PSU Rod Bundle Heat Transfer (RBHT) test facility","authors":"Grant R. Garrett ,&nbsp;Douglas J. Miller ,&nbsp;Turki Almudhhi ,&nbsp;Fan-Bill Cheung ,&nbsp;Brian R. Lowery ,&nbsp;Molly K. Hanson ,&nbsp;Stephen M. Bajorek ,&nbsp;Kirk Tien ,&nbsp;Chris L. Hoxie","doi":"10.1016/j.nucengdes.2024.113539","DOIUrl":"10.1016/j.nucengdes.2024.113539","url":null,"abstract":"<div><p>This paper evaluates the performance of the U.S. Nuclear Regulatory Commission’s (NRC’s) thermal hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) against experimental reflood data from the NRC/Pennsylvania State University (NRC/PSU) Rod Bundle Heat Transfer (RBHT) test facility, as an integral step in verification of code accuracy. This paper is an extension of the NURETH-20 conference paper by the first author (Garrett et al., 2023) that has been recommended for consideration and submission to Nuclear Engineering and Design. An international study on reflood thermal-hydraulics, sponsored by the Nuclear Energy Agency (NEA) Working Group on Accident Management and Analysis (WGAMA), was conducted with data collected in the NRC/PSU RBHT test facility, located at the Pennsylvania State University. A series of 16 benchmark tests were conducted, with conditions covering a carefully selected range of oscillatory, variable stepped and constant rate reflood injection velocities. These unique conditions are useful for code validation and model improvement. These 16 tests were segmented into 11 open tests, followed by five blind tests. This paper covers the five blind tests as the 11 open tests were covered by Garrett et al. at the NURETH-19 conference (Garrett et al., 2021).</p><p>For TRACE code benchmarking, a numerical model with the same dimensions as the RBHT facility was used. The initial and boundary conditions for this model were taken from experimental measurements. Many of the test conditions were chosen to examine sensitivities to important parameters, such as reflood liquid subcooling, reflood rate, and system pressure. The wide range of test conditions served to test the code and provide insight to its strengths and potential areas of improvement. These novel experiments were vital in this effort.</p><p>Simulations were made for five reflood tests and comparisons between predicted and measured results were made for the transient cladding temperatures, vapor temperature, bundle liquid mass fraction, carryover fraction, and steam exhaust fraction. The comparison presented in this paper has provided useful insight into code improvements. Studies to more accurately model reflood phenomena are currently underway as a result of the work presented in this paper.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142151754","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Fretting wear mechanism of nuclear fuel cladding tube under different tangential displacement","authors":"Jialing Li ,&nbsp;Huoming Shen ,&nbsp;Yehong Liao ,&nbsp;Yuxing Wang ,&nbsp;Songye Jin ,&nbsp;Zhenxun Peng ,&nbsp;Kaimo Wang ,&nbsp;Qisen Ren","doi":"10.1016/j.nucengdes.2024.113567","DOIUrl":"10.1016/j.nucengdes.2024.113567","url":null,"abstract":"<div><p>The fretting wear behavior of Zr-1Nb alloy cladding tubes under different tangential displacement amplitudes under grid-to-rod contact conditions was investigated. The dependence of the morphology of wear scars, microstructure, and wear mechanisms on the tangential displacement amplitudes was analyzed. The results indicate that as the tangential displacement amplitude increases, the fretting regime transitions from the mixed fretting regime to the gross slip regime gradually, moreover, the coefficient of friction initially increases and then decreases, while the wear volume and maximum wear depth gradually increase. In the mixed fretting regime, the primary wear mechanism is adhesive wear, whereas the primary wear mechanism in the gross slip regime is delamination. Moreover, the extent of oxidative wear becomes more severe while the tangential displacement amplitude is larger.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142135973","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}