loss of coolant accident
Recently Published Documents


TOTAL DOCUMENTS

692
(FIVE YEARS 122)

H-INDEX

16
(FIVE YEARS 3)

Author(s):  
Yogendra S. Garud ◽  
Andrew K. Hoffman ◽  
Raul B. Rebak

AbstractThe US Department of Energy is working with fuel vendors to develop accident tolerant fuels (ATF) for the current fleet of light water reactors (LWRs). The ATF should be more resilient to loss of coolant accident scenarios and help extending the life of the operating LWRs. One of the proposed ATF concepts is to use iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel. A concern in using ferritic FeCrAl is that this type of cladding may result in an increase in the concentration of tritium in the coolant. The objective of the current critical review is to collect and assess information from the literature regarding diffusion or permeation of hydrogen (H) and its isotopes deuterium (D) and Tritium (T) across industrial alloys (including FeCrAl) used or intended for the nuclear industry. Over a hundred years of data reviewed shows that the solubility of hydrogen in ferritic alloys is lower than in austenitic alloys but hydrogen permeates faster through a ferritic material than through austenitic materials. The tritium permeation rates in FeCrAl alloys are between those in austenitic stainless steels and in ferritic FeCr steels. The activation energy for hydrogen permeation is approximately 30 pct higher in the austenitic alloys compared with the ferritic (typically ∼ 50 kJ/mol in ferritic vs. ∼ 65 kJ/mol in the austenitic). None of the major elements in FeCrAl alloys react with hydrogen to form detrimental hydride phases. The effect of surface oxides on FeCrAl delaying hydrogen entrance into FeCrAl alloy is not part of this review.


2021 ◽  
Vol 23 (2) ◽  
Author(s):  
Putu Brahmanda Sudarsana ◽  
Wayan Nata Septiadi ◽  
Mulya Juarsa

SMART (System-Integrated Advanced Modular Reactor) merupakan desain reaktor multifungsi Generasi III+ tipe SMR (Small Modular Reactor) yang dikembangkan oleh KAERI (Korean Atomic Energy Research Institute) dengan kapabilitas produksi listrik 107 MWe dan energi termal 365 MWt. Sistem SMART meliputi berbagai fitur keselamatan untuk mengatasi LOCA (Loss of Coolant Accident) dan skenario kecelakaan lainnya. Salah satu dari fitur tersebut adalah Passive Residual Heat Removal System (PRHRS) atau sistem pembuang sisa panas pasif yang bekerja tanpa membutuhkan sumber daya elektrik. Sistem ini bekerja sesuai dengan prinsip sirkulasi alam sehingga bergantung pada aspek termal, tekanan, dan pengaruhnya terhadap aliran massa. Ketiga aspek tersebut dapat mempengaruhi kapabilitas pembuangan panas pada sistem. Data performa PRHRS reaktor SMART pada beberapa kondisi kecelakaan yang diperoleh melalui studi eksperimental maupun simulasi termohidrolika dianalisis pada kajian ini. Hasil analisis menunjukkan unjuk kerja pembuangan sisa panas yang baik oleh PRHRS SMART dengan waktu aktuasi yang tepat dan pendinginan yang stabil. Dengan kapabilitas multifungsi dan kemampuan pendinginan yang baik pada berbagai skenario kecelakaan, SMART memiliki potensi tinggi untuk kelak diterapkan di Indonesia.


2021 ◽  
Vol 13 (24) ◽  
pp. 14042
Author(s):  
Wei Sun ◽  
Chao Xu ◽  
Yi-Zhen Wang ◽  
Sui-Zheng Qiu ◽  
Yu-Sheng Liu ◽  
...  

Deterministic safety analysis (DSA) is essential for nuclear power plant licensing. The conservative method followed CFR50 Appendix K, which will lead to a large margin. As one of the DSA methodologies, best estimate plus uncertainty (BEPU) generates more realistic results that can be used in the license application of nuclear power plants (NPPs). However, uncertainty evaluation of parameters is needed in BEPU. In this article, the safety regulatory focuses on the large break loss of coolant accident (LBLOCA) of an advanced PWR. The BEPU analysis is mainly performed by TRACE V5.0 patch 4 code, and the uncertainty analysis is conducted based on DAKOTA code. For correlation coefficients analysis, the sample size is enlarged reasonably. According to the results, this NPP meets the acceptance criteria effectively in LBLOCA with enough margin. By statistic assessment, the set of PCTs calculated has typical normal distribution characters. Based on BEPU, the uncertainties of parameters are studied. Additionally, the influence of sample size on the correlation of parameters is considered too. It could be seen that more samples could permit a more accurate estimation for Spearman partial correlation coefficient (abbreviated as SPCC). The conclusions of this article can provide technical support for the subsequent review of the safety analysis report and the design changes of NPPs.


Energies ◽  
2021 ◽  
Vol 14 (24) ◽  
pp. 8527
Author(s):  
Marica Eboli ◽  
Francesco Galleni ◽  
Nicola Forgione ◽  
Nicolò Badodi ◽  
Antonio Cammi ◽  
...  

The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (water-cooled lead-lithium breeding blanket). Research activities are ongoing to master the phenomena and processes that occur during the postulated accident, to enhance the predictive capability and reliability of numerical tools, and to validate computer models, codes, and procedures for their applications. Following these objectives, ENEA designed and built the new separate effects test facility LIFUS5/Mod3. Two experimental campaigns (Series D and Series E) were executed by injecting water at high pressure into a pool of PbLi in WCLL-BB-relevant parameter ranges. The obtained experimental data were used to check the capabilities of the RELAP5 system code to reproduce the pressure transient of a water system, to validate the chemical model of PbLi/water reactions implemented in the modified version of SIMMER codes for fusion application, to investigate the dynamic effects of energy release on the structures, and to provide relevant feedback for the follow-up experimental campaigns. This work presents the experimental data and the numerical simulations of Test E4.1. The results of the test are presented and critically discussed. The code simulations highlight that SIMMER code is able to reproduce the phenomena connected to PbLi/water interaction, and the relevant test parameters are in agreement with the acquired experimental signals. Moreover, the results obtained by the first approach to SIMMER-RELAP5 code-coupling demonstrate its capability of and strength for predicting the transient scenario in complex geometries, considering multiple physical phenomena and minimizing the computational cost.


Author(s):  
Mohammed F Uddin ◽  
Gery Wilkowski ◽  
Sureshkumar Kalyanam ◽  
Frederick W. Brust

Abstract In typical leak-before-break (LBB) analyses in the nuclear industry, the uncracked piping normal operating forces and moments are applied in a cracked-pipe analytical procedure to determine normal leakage, and the combined forces and moments under normal operating condition and safe shutdown earthquake seismic loading are used in a fracture analysis to predict margins on "failure". The International Piping Integrity Research Program (IPIRG) performed in 1990 to 1998 provided some insights to typical LBB behaviors where pipe system tests were conducted with simulated seismic loadings. The test results showed a large margin on LBB which was also recognized in 2011 when the Argentinian Atucha II plant was analyzed using a robust full FE model. It was found that when circumferential through-wall cracks were put in the highest stressed locations, the applied moment dropped for both normal operating and N+SSE loading as the crack length increased. The through-wall crack size for causing a double ended guillotine break (DEGB) was greater than 90%-percent of the circumference. Similar results were also found for a petrochemical pipe system where thermal expansion stresses are much higher than the primary stresses. Even with very low toughness materials, the critical crack size leading to DEGB was greater than 80% of the circumference. The implication of this work is that pragmatically there is much higher margin for DEGB failure in nuclear plant operation, and efforts would be better focused on the potential for a small-break loss-of-coolant accident (SB-LOCA).


2021 ◽  
pp. 288-297
Author(s):  
Stepan Lys ◽  
◽  
Oksana Yurasova ◽  
Igor Galyanchuk

2021 ◽  
Vol 9 ◽  
Author(s):  
LI Xiao-ling ◽  
Wu Rong-jun ◽  
Xu Xiao-hui ◽  
Zhang Duo-fei ◽  
YU Ming

An optimization design and application of high temperature–resistant shielding material was carried out according to the nuclear power plant source characteristics and special protection requirements such as loss-of-coolant accident (LOCA). The composition of lead–boron polyethylene shielding composite was optimized based on the genetic algorithm and Monte Carlo methods and then realized by blending modification and graft copolymerization to improve its high temperature–resistant, shielding, and mechanical properties. Then comprehensive properties such as mechanical, neutron shielding, damp heat aging, irradiation resistance, and high temperature resistance were tested. These experiments proved that the high temperature–resistant lead–boron polyethylene shielding composite has excellent performance; especially, as it is able to keep a complete structure in a high-temperature environment of up to 190°C for 48 h. Finally, the shielding composite was applied to the shielding door design of a reactor pit chamber. When the shield thickness is 60 mm, the level of the neutron dose rate was reduced by 10 times, and that of the γ dose rate was reduced by 5 times, which meets all the requirements of radiation protection safety for nuclear power plants.


Sign in / Sign up

Export Citation Format

Share Document