Annals of Nuclear Energy最新文献_第10页

Validation and parametric study of FFRD model in DRACCAR code DRACCAR 代码中 FFRD 模型的验证和参数研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-08-05 DOI: 10.1016/j.anucene.2024.110832

{"title":"Validation and parametric study of FFRD model in DRACCAR code","authors":"","doi":"10.1016/j.anucene.2024.110832","DOIUrl":"10.1016/j.anucene.2024.110832","url":null,"abstract":"<div><p>The recent revision of Emergency Core Cooling System (ECCS) regulations in Korea has necessitated the consideration of Fuel Fragmentation, Relocation, and Dispersal (FFRD) phenomena in nuclear reactor safety analyses. Consequently, the Korea Atomic Energy Research Institute (KAERI) has developed and integrated the FFRD model into the domestically licensed safety analysis code, SPACE. Globally, US NRC’s FRAPTRAN and IRSN’s DRACCAR are available for evaluating FFRD phenomena. The FRAPTRAN code incorporates the FFR model, developed by Quantum Technology, while the fuel dispersal model is currently not included. In contrast, the DRACCAR code functions as an integrated analysis platform capable of modeling multi-dimensional thermal, hydraulic, mechanical, and chemical phenomena during a Loss of Coolant Accident (LOCA). This study conducts a thorough examination of the FFRD model in the DRACCAR code and validates its applicability through analysis using the Halden IFA-650 tests. The results demonstrate satisfactory predictive capabilities. Furthermore, a parametric study of key FFRD model parameters enhances the understanding of the FFRD model in the DRACCAR code. The development of detailed physical models in the FFRD model could significantly enhance the performance of the DRACCAR code, warranting the establishment of a comprehensive framework for these advancements. In the future, code-to-code comparisons between the DRACCAR code and other domestically developed integrated analysis platforms will be conducted to investigate various phenomena in depth.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141944532","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}

引用次数: 0

R2CA H2020 project for Reduction of Radiological Consequences of design basis and design extension Accidents R2CA H2020 减少设计基础和设计扩展事故的辐射后果项目

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-08-03 DOI: 10.1016/j.anucene.2024.110804

引用次数: 0

Experimental study on 100 hour-term performance of high-temperature sodium heat pipes 高温钠热管 100 小时性能实验研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-08-02 DOI: 10.1016/j.anucene.2024.110799

{"title":"Experimental study on 100 hour-term performance of high-temperature sodium heat pipes","authors":"","doi":"10.1016/j.anucene.2024.110799","DOIUrl":"10.1016/j.anucene.2024.110799","url":null,"abstract":"<div><p>High-temperature heat pipes, utilizing alkali metals as working fluids, are essential heat transfer components in cooling systems of nuclear reactors. To assess the long-term isothermal behavior and heat transfer performance variation of high-temperature sodium heat pipes, this research presents the design and construction of a long-duration experimental test rig for high-temperature heat pipes. A 100-hour experimental investigation was conducted under operating conditions of 900 °C. The results demonstrate that the heat pipes can operate stably for extended periods after startup. The average temperature at the condenser section gradually increased, while the overall temperature difference fluctuation decreased. The magnitude of temperature difference reduction was measured to be 2.3 °C, and the effective thermal resistance of the heat pipe decreased to 0.0639 K/W. These results suggest an enhancement in both the isothermal performance and heat transfer capability characteristics of the sodium heat pipe after long-term testing. This study provides valuable insights for the design and assessment of high-temperature heat pipe systems in nuclear reactor cooling applications.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141944534","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}

引用次数: 0

CORTEX experiments, Part III: Experimental determination of the zero power transfer function of AKR-2 with reliable uncertainties CORTEX 实验，第三部分：以可靠的不确定性对 AKR-2 的零功率传递函数进行实验测定

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-08-01 DOI: 10.1016/j.anucene.2024.110686

{"title":"CORTEX experiments, Part III: Experimental determination of the zero power transfer function of AKR-2 with reliable uncertainties","authors":"","doi":"10.1016/j.anucene.2024.110686","DOIUrl":"10.1016/j.anucene.2024.110686","url":null,"abstract":"<div><p>The transfer function is an important characteristic quantity of a nuclear reactor, since it contains the kinetic parameters. It expresses the response of a nuclear reactor to a disturbance of a certain frequency. If it is determined experimentally, it can be used to draw conclusions about the kinetic parameters. This article presents results of measurements of the zero power transfer function of the AKR-2 reactor at TU Dresden together with new data analysis methods. These measurements are compared to the theoretical zero power transfer function with kinetic parameters obtained via Monte Carlo simulations with MCNP and Serpent. To this end, advanced data analysis techniques based on a bootstrapping algorithms are employed. These techniques suppress the signal outside multiples of the fundamental frequency and additionally allow to obtain the full probability distribution of a peak in the frequency domain. This allowed for a reliable estimation of the mean value and uncertainty estimates of measured data of the zero power transfer function and the quantification of deviations between the experiments and the computations. It also made it possible to determine the phase of the zero power transfer function of AKR-2 for the first time. The experiments and computations are in agreement within the estimated uncertainties.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141962181","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}

引用次数: 0

Development and verification of multigroup advanced semi analytic nodal method solver for HTGR analysis with MHTGR-350 利用 MHTGR-350 开发和验证用于 HTGR 分析的多组高级半解析节点法求解器

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-08-01 DOI: 10.1016/j.anucene.2024.110818

{"title":"Development and verification of multigroup advanced semi analytic nodal method solver for HTGR analysis with MHTGR-350","authors":"","doi":"10.1016/j.anucene.2024.110818","DOIUrl":"10.1016/j.anucene.2024.110818","url":null,"abstract":"<div><p>This study introduces the multigroup advanced semi-analytic nodal method (A-SANM) tailored for the high-temperature gas-cooled reactor (HTGR) analysis. The A-SANM has been crafted specifically for reactors with hexagonal geometries, such as the Vodo-Vodyanoi energetichesky reactor (VVER) and HTGR. A triangular node was constructed with a 12-term basis to delineate the flux by integrating both the polynomial and hyperbolic functions. The multigroup calculation kernel of this approach was embedded in the nodal diffusion code, RAST-V. To evaluate the computational efficiency of the A-SANM, we employed the MHTGR-350 benchmark. This benchmark, associated with a modular high-temperature gas-cooled reactor, was established by the OECD/NEA under the NGNP Project in 2021. In this study, we conducted the Phase I calculations to evaluate the performance of the neutronics code. Key parameters including the multiplication factor, rod worth, and axial and radial power distributions were meticulously assessed. When juxtaposed with the Monte Carlo code MCS, the A-SANM exhibited a deviation of –97 pcm. Differences in the axial and radial power were ± 4 and ± 3 %, respectively. Furthermore, the rod worth discrepancy was –6 pcm when set against the MCS. In summary, this study effectively elucidates the potential and precision of the multigroup A-SANM for the HTGR evaluations.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141944535","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}

引用次数: 0

Study on the specifications of the basic core configurations of the modified STACY 关于改进型 STACY 基本核心配置规格的研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-08-01 DOI: 10.1016/j.anucene.2024.110783

{"title":"Study on the specifications of the basic core configurations of the modified STACY","authors":"","doi":"10.1016/j.anucene.2024.110783","DOIUrl":"10.1016/j.anucene.2024.110783","url":null,"abstract":"<div><p>Since the compositions and properties of the fuel debris are uncertain, critical experiments are required to validate calculation codes and nuclear data used for the safety evaluation. For this purpose, the Japan Atomic Energy Agency (JAEA) has been modifying a critical assembly called “STACY.” The first criticality of the modified STACY is scheduled for spring 2024. This paper reports the consideration results of the core configurations of the modified STACY for the first criticality and inspections. The specifications of these core configurations were determined in advance, we tried to make them with a simplified computational model that considers the reactivity effect around the core. At the first criticality, two types of grid plates with different neutron moderation conditions (their hole spacings are 1.50 cm and 1.27 cm) were prepared. On the other hand, there is a limitation on the number of available UO<sub>2</sub> fuel rods. The core configurations for the first criticality satisfying these experimental constraints were designed by computational analysis. A cylindrical core configuration with a 1.50 cm pitch grid plate close to the optimum moderation condition needs 253 fuel rods to reach criticality. As to the 1.27 cm grid plate, we considered core configurations with 2.54 cm intervals by using doubled pitches of the grid plate. It will need 213 fuel rods for the criticality to be reached. In addition, the experimental core configuration was considered with steel/concrete simulant rods to simulate fuel debris conditions. This paper shows six core configurations with different conditions, and all of them satisfy the regulatory requests.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141944536","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}

引用次数: 0

Simulation study on the transient operating characteristics of equal height difference passive containment cooling system 等高差被动安全壳冷却系统瞬态运行特性模拟研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-07-30 DOI: 10.1016/j.anucene.2024.110808

{"title":"Simulation study on the transient operating characteristics of equal height difference passive containment cooling system","authors":"","doi":"10.1016/j.anucene.2024.110808","DOIUrl":"10.1016/j.anucene.2024.110808","url":null,"abstract":"<div><p>The passive containment cooling system is crucial important for marine nuclear power platform. This paper focuses on the transient characteristic of passive containment cooling system with equal height difference natural circulation under two starting strategy. Dynamic simulation model of passive containment cooling system is set up with APROS software. The simulation results indicate that the initial containment pressure, initial non-condensable gas content, and inlet water temperature have important effects on the pressure drop rate and natural circulation flow rate. And obvious differences exist between the empty tube starting mode and the full tube starting mode. Under single-phase natural circulation conditions, the pressure drop rate of the empty tube starting mode is more than 3.1 times that of the full tube starting mode, while under two-phase natural circulation conditions, the pressure drop rate of the empty tube starting mode is more than 1.5 times that of the full tube starting mode, since large fluctuation amplitude of two-phase flow instability caused by condensation induced water hammer in the natural circulation occurs under full tube starting mode.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141862820","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}

引用次数: 0

Research on influence of wear ring clearance on energy loss of reactor coolant pump 磨损环间隙对反应堆冷却剂泵能量损失的影响研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-07-30 DOI: 10.1016/j.anucene.2024.110819

{"title":"Research on influence of wear ring clearance on energy loss of reactor coolant pump","authors":"","doi":"10.1016/j.anucene.2024.110819","DOIUrl":"10.1016/j.anucene.2024.110819","url":null,"abstract":"<div><p>Reactor coolant pump (RCP), the only rotating equipment in the nuclear island, directly determines the safety of the nuclear power system, has been likened to the “heart” of a nuclear power plant. The RCP needs to have very high reliability and stability. Its independent design and construction have always been the focus and difficulty of promoting the independent construction of nuclear power. To study the effect of the wear ring clearance on the energy loss characteristics of the RCP, the numerical simulation of the internal flow field of computation models with different wear ring clearance dimensions were conducted. The results indicate: As the wear ring clearance increases, the trend of performance reduction becomes more significant, especially under the large flow conditions; The increase of the wear ring clearance results in an increase of the pressure loss in the impeller and diffuser, and the variation in pressure distribution in different flow passages under different wear ring clearance dimensions is influenced by the position of the blade; The distribution of turbulent dissipation entropy production rate corresponds significantly to the distribution of direct dissipation entropy production rate, and the dissipation entropy production being more significantly affected by the wear ring clearance dimension closer to the shroud; The change of wear ring clearance dimension has limited impact on the wall shear stress entropy production of the impeller blade, while its effect on the diffuser blade is more pronounced; With the increase of the wear ring clearance, the scale of vortices near the impeller front shroud significantly increases, and the entropy production on the surface of vortex core at the impeller inlet gradually increases. This study is of great significance to ensure the secure and stable operation of the RCP, and provides an important reference for the hydraulic optimization design of the RCP.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141862816","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}

引用次数: 0

Modeling and flux reconstruction of unstructured geometric reactor 非结构化几何反应器的建模和通量重建

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-07-30 DOI: 10.1016/j.anucene.2024.110815

引用次数: 0

A Hankel Transform approach to Doppler broadening 多普勒增宽的汉克尔变换方法

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-07-30 DOI: 10.1016/j.anucene.2024.110790

引用次数: 0