International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第8页

Numerical study on PWR fuel rod cladding ballooning and burst behavior with the thermo-mechanical coupling finite element method 采用热力耦合有限元法对压水堆燃料棒包壳气胀爆行为进行数值研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 Epub Date: 2025-04-16 DOI: 10.1016/j.jandt.2025.04.003

Xin Zhang , Gen Li , Runhua Gao , Xinhai Zhao , Peitao Yao

{"title":"Numerical study on PWR fuel rod cladding ballooning and burst behavior with the thermo-mechanical coupling finite element method","authors":"Xin Zhang , Gen Li , Runhua Gao , Xinhai Zhao , Peitao Yao","doi":"10.1016/j.jandt.2025.04.003","DOIUrl":"10.1016/j.jandt.2025.04.003","url":null,"abstract":"<div><div>The ballooning behavior of fuel rod cladding is an important issue of nuclear reactor, which may initiate a severe nuclear reactor accident. Based on COMSOL multi-physical module, this research established a thermo-mechanical coupling finite element model, which can accurately simulate the distribution and evolution of temperature, strain, and stress of the fuel cladding under high temperature and pressure. The validation results indicated that the maximum deviations of the predicted temperature and strain are 8 % and 11 %, respectively, in comparison to the experimental data. Moreover, the burst data predicted by the model were consistent with the results of Yadav experiment. These tests encompassed a heating rate of 2 K/s to 8 K/s and an internal overpressure range of 1 MPa–9 MPa. The burst temperature predicted by the model decreased with the increase of internal overpressure, from 1315 K to approximately 1014 K. This trend was consistent with that observed in the experimental data, thereby verifying the model's accuracy and reliability. The effects of key parameters such as cladding heating rate and cladding internal pressure on cladding ballooning and burst behavior were further analyzed. The results indicated that as the heating power increased, the cladding burst temperature rose from 1055 K to 1290 K, while the burst strain decreased from 68 % to 27 %. When the internal overpressure rose from 3 MPa to 7 MPa, the burst strain and stress increased by 93 % and 174 %, respectively. This study elucidated thermo-mechanical cladding response, facilitating predictive safety analysis and optimal fuel rod design to mitigate loss-of-coolant accident risks in nuclear reactors.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 7-18"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143859354","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

Thermomechanical and radiological investigations of storage casks for radioactive spent fuel assemblies 放射性乏燃料组件贮存桶的热机械和放射学研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 Epub Date: 2025-04-23 DOI: 10.1016/j.jandt.2025.04.005

M. Mehdizadeh, M. Aghaie, M. Sharifi

{"title":"Thermomechanical and radiological investigations of storage casks for radioactive spent fuel assemblies","authors":"M. Mehdizadeh, M. Aghaie, M. Sharifi","doi":"10.1016/j.jandt.2025.04.005","DOIUrl":"10.1016/j.jandt.2025.04.005","url":null,"abstract":"<div><div>In this research, the Monte Carlo particle transport method and COMSOL Multiphysics were utilized to simulate and analyze the TK-13 cask, which contains spent fuel assemblies of the WWER-1000 reactor. Initially, the cask was modeled as being filled with water, and its criticality was assessed by evaluating the effective multiplication factor (Keff) while considering various neutron shielding materials, including polyethylene, borated polyethylene, ethylene glycol, and water solutions. The Keff value was calculated by varying the boron percentage in the holding basket of spent fuel assemblies and adjusting the pitch between these assemblies. Additionally, the total neutron and gamma dose rates on the cask's outer wall were measured. A thick steel wall, 34 cm in thickness, was modeled around the spent fuel assemblies, and a concrete wall was also considered, with total dose rates calculated for both cases. The uncertainty in these measurements was also quantified. The cask was further simulated in COMSOL Multiphysics to evaluate thermal and mechanical parameters. Thermal analysis was conducted in both steady and unsteady states; for steady-state analysis, the minimum and maximum fuel burnup scenarios were examined, along with the effects of fins on the outer wall. In the unsteady thermal analysis, temperature variations at different points within the cask and their distribution over a hundred-year period were investigated. The mechanical analysis focused on Von Mises stress and displacement values at various locations within the cask. Finally, the results were compared with findings from other studies, leading to comprehensive conclusions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 29-42"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143877037","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

Preliminary study on the postulated siting accident source term of integrated small reactor 关于综合小型反应堆选址事故源假设条件的初步研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 Epub Date: 2025-04-18 DOI: 10.1016/j.jandt.2025.04.006

Dawei Sun, Shengqin Gao, Liang Zhai, Qiliang Mei, Yaru Fu, Jie Mao

{"title":"Preliminary study on the postulated siting accident source term of integrated small reactor","authors":"Dawei Sun, Shengqin Gao, Liang Zhai, Qiliang Mei, Yaru Fu, Jie Mao","doi":"10.1016/j.jandt.2025.04.006","DOIUrl":"10.1016/j.jandt.2025.04.006","url":null,"abstract":"<div><div>In accordance with the characteristics of integrated small reactor, a study was conducted on the postulated siting accident source term scheme. The principles for selecting accident types were explored for both the design basis accident condition and the design extension condition. For the deterministic design basis accident condition, a representative accident, such as a fuel handling accident, was identified from the perspective of envelope; for the design extension condition, a representative accident was identified based on the likelihood of accident occurrence, such as when the screening frequency is 10<sup>−7</sup>/(reactor·year). When the probability of core damage is extremely low, it is considered to construct at least one entire core fuel cladding damage accident as a representative case. Focusing on the radioactive production, migration, removal, and release, a study was conducted on the source term analysis models and key parameter values of various representative accidents, thus simulating the radioactive activity released into the environment by the accident and identifying <sup>131</sup>I as the dominant nuclide that should be given special attention. Taking a coastal plant site as the object, the dose to the public staying at 16 locations along the site boundary after the accident was further evaluated, and it was demonstrated that even reducing the site boundary to 100 m around the reactor can meet the dose limit requirements of the “small reactor review principle”, and the suitability of the source term schemes was certificated.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 1-6"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143859353","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

The impact of natural removal on the release of nuclides into the containment after accident 事故发生后，自然移除对核素释放到安全壳的影响

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 Epub Date: 2025-04-15 DOI: 10.1016/j.jandt.2025.04.002

Haiying Chen, Xinli Gao, Pingfei Du, Yuan Gao, Fudong Liu

{"title":"The impact of natural removal on the release of nuclides into the containment after accident","authors":"Haiying Chen, Xinli Gao, Pingfei Du, Yuan Gao, Fudong Liu","doi":"10.1016/j.jandt.2025.04.002","DOIUrl":"10.1016/j.jandt.2025.04.002","url":null,"abstract":"<div><div>The natural removal coefficient directly affects the amounts of radioactive iodine and aerosols released into the containment after the accident. The radioactivity calculation models in LOCA were established, and the impact of natural removal coefficients on the release of elemental iodine and aerosols into containment was quantitatively analyzed. The results showed that due to the comprehensive influence of factors such as core release type, release time, natural removal effect, nuclide decay, and containment leakage, the radioactivity of each nuclide in the containment reached its maximum value after 40 min of the accident, and then gradually decreased over time. During the effective natural removal time, there was a significant difference in the radioactivity of elemental iodine in the containment under different natural removal coefficients. Taking <sup>131</sup>I as an example, the radioactivity ratio of elemental <sup>131</sup>I in containment corresponding to the two natural removal coefficients decreased first and then increased over time. Finally, the radioactivity of elemental <sup>131</sup>I under different natural removal coefficients was basically the same. The change of aerosol radioactivity in the containment was obviously affected by the value of natural removal coefficients. Under two different natural removal coefficients, the maximum radioactivity ratio of the aerosol nuclides in the containment was about 2.3.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 43-47"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143891342","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

Debris filtering efficiency assessment of the fuel assembly bottom nozzle 燃油组件底部喷嘴碎屑过滤效率评估

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 Epub Date: 2025-04-22 DOI: 10.1016/j.jandt.2025.04.004

Zheng Mei-yin, Jiao Yong-jun, Li Zheng-yang, Ren Quan-yao, Xiao Zhong, Su Min

{"title":"Debris filtering efficiency assessment of the fuel assembly bottom nozzle","authors":"Zheng Mei-yin, Jiao Yong-jun, Li Zheng-yang, Ren Quan-yao, Xiao Zhong, Su Min","doi":"10.1016/j.jandt.2025.04.004","DOIUrl":"10.1016/j.jandt.2025.04.004","url":null,"abstract":"<div><div>The debris-induced fretting wear is a key factor to cause the fuel failure in pressurized water reactor (PWR). Assessing the debris filtering efficiency of bottom nozzle is an important way to analyze the reliability of fuel assembly. The debris filtering efficiency assessment method of bottom nozzle was proposed, which concludes experimental research method and numerical calculation method. The experimental method concludes A-types debris (ball, hex nut, spring, gasket, and pin) and B-types debris (cylindrical metal wire, helical metal cutting, and metal strip). The experimental method demonstrates that the bottom nozzle with hemispherical surface structure presents the filtering efficiency of 100 % for B-types debris. The numerical simulation method is a coupling process of CFD and DEM, which mainly analyze the movement and collision of debris. The bottom nozzle shows better filtering efficiency for cylindrical debris, and the filtering efficiency in the numerical simulation is in good accordance with the experimental results. The filtering efficiency assessment method is suitable for evaluating the debris filtering behavior of bottom nozzle and helpful for the design and verification of bottom nozzle.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 48-54"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143899108","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

Analysis of natural circulation characteristics for pool-type fast reactor system 池式快堆系统自然循环特性分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 Epub Date: 2025-04-17 DOI: 10.1016/j.jandt.2025.04.007

Yufeng Lv, Zhiwei Zhou, Xingmin Liu

{"title":"Analysis of natural circulation characteristics for pool-type fast reactor system","authors":"Yufeng Lv, Zhiwei Zhou, Xingmin Liu","doi":"10.1016/j.jandt.2025.04.007","DOIUrl":"10.1016/j.jandt.2025.04.007","url":null,"abstract":"<div><div>In the inherent safety design of the pool-type fast reactor, when the normal heat transfer routine of the primary system is lost during accident conditions, the residual heat of the reactor core is transferred to the ultimate heat sink through the natural circulation in the reactor vessel. The natural circulation phenomenon in the pool-type reactor system under accident transient is complicated. For the purpose to improve the knowledge of natural circulation characteristics of the pool-type reactor system and validate the applicability and accuracy of system analysis code for simulating natural circulation in pool systems, the RELAP5 MOD3.2 code was adopted to calculate the station blackout (SBO) test conducted on a pool-type test facility. The analysis results show that in the early stage of the transient process, the flow re-distribution phenomenon occurs in the pool-type vessel. There exists natural circulation flow path both inside core assembly simulators and outside core assembly simulators. Due to the influence of the hot fluid at the outlet of the core and the cold fluid at the outlet of the DHX, the thermal stratification phenomenon occurs in hot pool. The main thermal-hydraulic parameters calculated by the code (such as the fluid outlet temperature in the core assembly simulator, the flow rate of the primary circuit and the fluid inlet and outlet temperature of the DHX primary side) are in good agreement with the experimental results. Through the conduct of this work, the feasibility of the RELAP5 code calculation model and modeling method is verified, which can be further adopted to the natural circulation simualtion of pool-type reactors.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 19-28"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143865130","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

An open-source Thermo-hydraulic Uniphase Advection and Convection Solver for Salt Flows (TUAS) 开源热液单相平流对流求解器（TUAS）

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2024-12-01 Epub Date: 2025-03-28 DOI: 10.1016/j.jandt.2025.03.006

Theodore Kay Chen Ong , Sicong Xiao , Per F. Peterson

{"title":"An open-source Thermo-hydraulic Uniphase Advection and Convection Solver for Salt Flows (TUAS)","authors":"Theodore Kay Chen Ong , Sicong Xiao , Per F. Peterson","doi":"10.1016/j.jandt.2025.03.006","DOIUrl":"10.1016/j.jandt.2025.03.006","url":null,"abstract":"<div><div>Due to a lack of open-source GUI simulators and corresponding system level simulation libraries for Gen IV Reactors such as the Fluoride Salt Cooled High Temperature Reactor (FHR), a systems level library for single phase flow named “TUAS” was written in the Rust which was then used to construct the first educational simulator of the Compact Integral Effects Test (CIET). In this work, we successfully performed a verification and validation (V&V) study on TUAS using idealized natural circulation simulation results from the System Analysis Module (SAM) simulation of CIET, experimental data of a HITEC salt to YD-325 oil Shell and Tube Heat Exchanger (STHE) as well as experimental forced circulation data from CIET. TUAS was shown to match SAM results well for the natural circulation verification study and experimental data well for forced circulation validation studies. Real-time simulation capabilities on consumer laptops for forced circulation were also demonstrated for the Open Source TUAS library using the CIET Educational Simulator with potential for validated real-time natural circulation in future work. Moreover, TUAS could be coupled with multiphase solvers and neutronics feedback in future to build real-time 2D and 3D reactor simulators for both research and outreach. These efforts contribute to a richer open-source nuclear reactor safety simulation ecosystem built in modern system programming languages such as Rust. Such an ecosystem would provide tools for a more conducive environment in education and research for nuclear engineering.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 4","pages":"Pages 281-301"},"PeriodicalIF":0.0,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145860888","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

Study of a multi-factor coupling removal model for gas-borne source terms under jet flow conditions 射流条件下气源项多因素耦合去除模型研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2024-12-01 Epub Date: 2025-03-15 DOI: 10.1016/j.jandt.2025.03.003

Chao Tian , Junyan Chen , Shisi Cheng , Dongchuan Ying , Zian Zhai , Lili Tong , Qiang Lv

{"title":"Study of a multi-factor coupling removal model for gas-borne source terms under jet flow conditions","authors":"Chao Tian , Junyan Chen , Shisi Cheng , Dongchuan Ying , Zian Zhai , Lili Tong , Qiang Lv","doi":"10.1016/j.jandt.2025.03.003","DOIUrl":"10.1016/j.jandt.2025.03.003","url":null,"abstract":"<div><div>After a pipe rupture in the reactor steam generator, high-temperature and high-pressure radioactive materials from the primary circuit enter the secondary side of the steam generator via a jet flow. During this process, the physical properties of the water in the liquid pool change as the washing progresses, which in turn affects the removal of radioactive materials. To investigate the multi-factor coupling removal characteristics of aerosols under jet flow conditions, a study was conducted on a jet flow aerosol washing removal model. By integrating a thermal-hydraulic model, an analysis program suitable for jet flow conditions was developed and validated. The validation process demonstrated that the developed model can accurately predict the jet flow washing process of aerosols. The study using this model revealed that as the submergence depth increases, the DF in the rising zone gradually increases, while the DF in the injection zone gradually decreases. Even at higher submergence depths, the jet flow aerosol removal in the injection zone continues to make a significant contribution to the overall washing process. Under the same submergence depth conditions, as the mass flux increases, the contribution of the DF in the injection zone becomes more pronounced, significantly influencing the jet flow aerosol removal mechanism during the washing process. When comparing the results of the developed jet flow washing removal model with those of existing international aerosol washing analysis programs, it was found that the developed model provides more accurate overall predictions of the DF for aerosol washing.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 4","pages":"Pages 239-247"},"PeriodicalIF":0.0,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145860890","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

24-Month long cycle fuel management research based on accident tolerant fuel 基于事故容错燃料的24月长周期燃料管理研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2024-12-01 Epub Date: 2025-03-28 DOI: 10.1016/j.jandt.2025.03.005

Hong Cao, Chen Ye, Wenhua Zhang, Lihua Wang, Qingxiang Yang

{"title":"24-Month long cycle fuel management research based on accident tolerant fuel","authors":"Hong Cao, Chen Ye, Wenhua Zhang, Lihua Wang, Qingxiang Yang","doi":"10.1016/j.jandt.2025.03.005","DOIUrl":"10.1016/j.jandt.2025.03.005","url":null,"abstract":"<div><div>As the installed nuclear power plants steadily increase in China, the demand for 24-month long cycle fuel management is gradually increasing as well. 18-month refueling strategy is mainly applied in PWRs currently, and the fuel enrichment and burnup have already reached the limits. By utilizing the accident tolerant fuel (ATF), fuel performance can support higher rod burnup and fuel enrichment, a more economical 24 month refueling cycle strategy can be achieved. In this paper, with the enrichment and burnup limits improved based on ATF fuel, 24-month fuel management strategies are designed. The core design meets the requirements of safety related acceptance criteria. Fuel rod performance analysis results indicate that, to achieve a higher burnup, it is necessary to introduce coated cladding and large grain pellet to mitigate cladding corrosion and fission gas release. This study provides technical support for the longer cycle operation of passive nuclear power plants and points out the direction for improving fuel enrichment and rod burnup limits.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 4","pages":"Pages 248-255"},"PeriodicalIF":0.0,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145860891","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

Review of spatial scale dispersion models (ATDMs) to simulate environmental dispersion and deposition of radionuclides and the overview of GIS coupling with dispersion models 模拟放射性核素环境弥散和沉积的空间尺度弥散模型综述及GIS与弥散模型耦合的研究进展

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2024-12-01 Epub Date: 2025-03-14 DOI: 10.1016/j.jandt.2025.03.004

Hussaini Muhammad , Wang Xuan , Mingjun Wang , Guanghui Su

{"title":"Review of spatial scale dispersion models (ATDMs) to simulate environmental dispersion and deposition of radionuclides and the overview of GIS coupling with dispersion models","authors":"Hussaini Muhammad , Wang Xuan , Mingjun Wang , Guanghui Su","doi":"10.1016/j.jandt.2025.03.004","DOIUrl":"10.1016/j.jandt.2025.03.004","url":null,"abstract":"<div><div>Estimation of the radiological consequences including the calculation of radiation doses to the exposed population requires numerical evaluation of atmospheric radioactive releases from nuclear facilities under accidental scenarios for implementation of adequate protective and other response actions to prevent or minimize severe deterministic effects and reduce the probability of occurrence of stochastic effects. This paper reviews various state-of-the-art spatial scale atmospheric dispersion models, known as the ATDMs, designed to evaluate atmospheric transport, dispersion, and depositions of radionuclide emissions from release sources. The ATDMs (atmospheric transport, dispersion, and deposition models) reviewed in this paper are FLEXPART, CALPUFF, LADAS, RASCAL, RIMPUFF, and HotSpot. These ATDMs are increasingly applied to perform probabilistic radiological risk assessment, which is one of the IAEA safety requirements for site evaluation of new NPP, and the calculated doses can be used to establish EPZs, an underlying basis of emergency preparedness. The numerical approaches, modeling methodology, advantages, limitations, scope of applications, and research investigations performed using these models in radiological and other environmental impact assessments within their scope of applications are presented. GIS has been widely recognized as a technique applied to aid the modeling of pollutant dispersion and visualization of model outputs in a spatial context, among other benefits. In nuclear accident consequence evaluation, spatial data provides key information such as accident location, range of radiological contamination, population density, etc, which are very important. This demonstrates the high significance of GIS-ATDMs coupling. Therefore, an overview of the coupling of dispersion models with GIS is also highlighted.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 4","pages":"Pages 256-280"},"PeriodicalIF":0.0,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145860889","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