Nuclear Engineering and Design最新文献

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Practical approach to Hazard-Consistent fragility curve estimates using Bayesian updating

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-28 DOI: 10.1016/j.nucengdes.2025.114029

Nevena Šipčić , Pablo García de Quevedo Iñarritu , Mohsen Kohrangi , Dimitrios Vamvatsikos , Paolo Bazzurro

Seismic fragility curves provide the probability of exceedance of a given damage state, should different levels of ground motion intensity be experienced at the site where the structure, or component, is located. Such curves are often derived via multiple nonlinear response history analyses (NLRHA) using sets of “suitable” ground motions that, in line with the best practice, should be consistent with the seismic hazard at the site. Based on the selected sets of records, one can estimate fragility functions that are often assumed to follow a lognormal distribution defined by two parameters, i.e., the logarithmic mean (µ) and the logarithmic standard deviation (β). Our focus is on estimating them using a state-of-the-art approach that involves hazard-consistent record selection via Conditional Spectrum and multiple stripe analysis. However, this approach usually requires many NLRHAs, with high computational costs, especially for the complex structural models typical of the nuclear industry. This study investigates the optimal number of ground motions and intensity levels required to keep the computational burden acceptable without compromising accuracy. To do so, we adopt a Bayesian framework with Markov chain Monte Carlo simulation and Metropolis–Hasting sampling. Our findings show that this approach effectively helps analysts best allocate computational resources while ensuring acceptable accuracy in estimating the probability of reaching or exceeding the considered damage states.

{"title":"Practical approach to Hazard-Consistent fragility curve estimates using Bayesian updating","authors":"Nevena Šipčić , Pablo García de Quevedo Iñarritu , Mohsen Kohrangi , Dimitrios Vamvatsikos , Paolo Bazzurro","doi":"10.1016/j.nucengdes.2025.114029","DOIUrl":"10.1016/j.nucengdes.2025.114029","url":null,"abstract":"<div><div>Seismic fragility curves provide the probability of exceedance of a given damage state, should different levels of ground motion intensity be experienced at the site where the structure, or component, is located. Such curves are often derived via multiple nonlinear response history analyses (NLRHA) using sets of “suitable” ground motions that, in line with the best practice, should be consistent with the seismic hazard at the site. Based on the selected sets of records, one can estimate fragility functions that are often assumed to follow a lognormal distribution defined by two parameters, i.e., the logarithmic mean (µ) and the logarithmic standard deviation (β). Our focus is on estimating them using a state-of-the-art approach that involves hazard-consistent record selection via Conditional Spectrum and multiple stripe analysis. However, this approach usually requires many NLRHAs, with high computational costs, especially for the complex structural models typical of the nuclear industry. This study investigates the optimal number of ground motions and intensity levels required to keep the computational burden acceptable without compromising accuracy. To do so, we adopt a Bayesian framework with Markov chain Monte Carlo simulation and Metropolis–Hasting sampling. Our findings show that this approach effectively helps analysts best allocate computational resources while ensuring acceptable accuracy in estimating the probability of reaching or exceeding the considered damage states.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114029"},"PeriodicalIF":1.9,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143715479","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

keff behaviour of a potential PWR assembly loaded disposal canister in simplified corrosion scenarios

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-27 DOI: 10.1016/j.nucengdes.2025.114007

M. Frankl , L. Berry , A. Vasiliev , D. Rochman , H. Ferroukhi , N. Diomidis , M. Wittel

The Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) and the Center for Nuclear Engineering and Sciences at the Paul Scherrer Institute (PSI) are investigating how potential long-term changes to the geometry and material composition of the current final disposal canister (BE-ELB) design may affect the neutron multiplication factor (k_eff) of a canister loaded with PWR spent nuclear fuel assemblies (FAs). Several conservative corrosion scenarios were formulated, modeled, simulated and analyzed using the Monte-Carlo codes MCNP6.2® and Serpent2.2. The corrosion-induced effects, such as the replacement of moderator by magnetite, differ significantly from typical fuel lattice configurations used in reactor or storage pool safety analyses. This paper therefore focuses on underlying physical phenomena causing the observed k_eff changes, including analyses of neutron currents and spectra, the ‘6-factor formula’, and the sensitivity of k_eff to specific regions, materials, and nuclides. These analyses showed the canister wall and the corrosion product magnetite to act as a heavy reflector, enhancing the backscattering of neutrons in the epithermal and fast energy ranges. Furthermore, the critical role of the water distribution in all the potential scenarios was revealed. Water inside the FAs clearly increases reactivity by moderation, water outside the remainders of the steel basket, however, has an inhibiting effect on neutron multiplication. All simulation results heavily depend on the specific preliminary ELB design. To that end, the results of this study can help to optimize the ELB design and the Swiss concept for the final disposal of high-level radioactive waste.

{"title":"keff behaviour of a potential PWR assembly loaded disposal canister in simplified corrosion scenarios","authors":"M. Frankl , L. Berry , A. Vasiliev , D. Rochman , H. Ferroukhi , N. Diomidis , M. Wittel","doi":"10.1016/j.nucengdes.2025.114007","DOIUrl":"10.1016/j.nucengdes.2025.114007","url":null,"abstract":"<div><div>The Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) and the Center for Nuclear Engineering and Sciences at the Paul Scherrer Institute (PSI) are investigating how potential long-term changes to the geometry and material composition of the current final disposal canister (BE-ELB) design may affect the neutron multiplication factor (<em>k<sub>eff</sub></em>) of a canister loaded with PWR spent nuclear fuel assemblies (FAs). Several conservative corrosion scenarios were formulated, modeled, simulated and analyzed using the Monte-Carlo codes MCNP6.2® and Serpent2.2. The corrosion-induced effects, such as the replacement of moderator by magnetite, differ significantly from typical fuel lattice configurations used in reactor or storage pool safety analyses. This paper therefore focuses on underlying physical phenomena causing the observed <em>k<sub>eff</sub></em> changes, including analyses of neutron currents and spectra, the ‘6-factor formula’, and the sensitivity of <em>k<sub>eff</sub></em> to specific regions, materials, and nuclides. These analyses showed the canister wall and the corrosion product magnetite to act as a heavy reflector, enhancing the backscattering of neutrons in the epithermal and fast energy ranges. Furthermore, the critical role of the water distribution in all the potential scenarios was revealed. Water inside the FAs clearly increases reactivity by moderation, water outside the remainders of the steel basket, however, has an inhibiting effect on neutron multiplication. All simulation results heavily depend on the specific preliminary ELB design. To that end, the results of this study can help to optimize the ELB design and the Swiss concept for the final disposal of high-level radioactive waste.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114007"},"PeriodicalIF":1.9,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706421","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}

引用次数: 0

Coupled analysis system development on heat pipe reactor

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-27 DOI: 10.1016/j.nucengdes.2025.114003

Sung Nam Lee, Sung Hoon Choi, Nam-il Tak, Hong-sik Lim, Chan Soo Kim

This study outlines the design tool and coupled analysis for heat transport in a space heat pipe reactor to provide power in a space environment for 10 years without fuel replacement. Korea Atomic Energy Research Institute (KAERI) has developed the design analysis tools to investigate the temperature distribution and maximum temperature in the reactor core. The neutronics code, McCARD, provides the power profile of the fuel compact in the core. The heat transport code, HEPITOS, predicts the temperature profiles coupled with the heat pipe analysis code, LUHPIS.

Since each code has different physics, the numerical calculation can be performed by the explicit method or the coupled method. Most studies have been analyzed using the coupled calculations of HEPITOS and LUHPIS. The coupled analysis with the neutronic code has been done to find out the feasibility of the coupled system development. The calculated results will be a reference for the evaluation of the thermal margins of the components.

引用次数: 0

CHF enhancement in downward-facing boiling surface using shrouds for calandria vessel during severe accident in PHWRs

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-26 DOI: 10.1016/j.nucengdes.2025.114010

P.K. Verma , P.P. Kulkarni , A.K. Nayak

The orientation of the heated surface significantly affects the boiling process. Boiling on a downward-facing surface is particularly challenging because bubble detachment is hindered, leading to longer bubble residence times and unique interactions than on vertical or inclined surfaces. This study investigates boiling on a large downward-facing flat surface (100 × 400 mm), focusing on critical heat flux (CHF) phenomenon. During the postulated severe accident, the situation arises in Pressurised Heavy Water Reactors (PHWRs) due to multiple failures of cooling systems and safety systems. The pressure tubes and calandria tubes have the potential to break, resulting in hot debris that falls to the bottom of the calandria vessel. The calandria vessel has a large curvature due to its larger diameter, and the bottommost portion is like a flat plate. To contain the hot debris or molten corium inside the vessel and maintain the integrity of the calandria vessel at a higher temperature is crucial to arrest the progress of a severe accident. The cooling of the vessel from outside without occurring CHF at the bottom location is important. Historically, downward-facing boiling has received limited attention, as it is normally not used in industrial applications owing to lower heat transfer and CHF values due to adverse buoyancy. Nonetheless, it is important to investigate because of the severe accident situation in PHWRs. Incorporating a simple technique of shrouds surrounding the calandria vessel can enhance the CHF by enhancing the buoyancy. This paper investigates the potential enhancement of CHF through the use of shrouds.

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引用次数: 0

Impact of γ irradiation and thermal aging on the swelling pressure of GMZ bentonite

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-26 DOI: 10.1016/j.nucengdes.2025.114012

Wei Liu , Dong Liang , Zhongtian Yang , Chao Gao , Jingli Xie , Meilan Jia

To investigate the combined effects of γ irradiation and thermal aging on the swelling pressure of buffer materials used in the geological disposal of high-level radioactive waste, Gaomiaozi (GMZ) bentonite from Inner Mongolia was herein selected. Samples were irradiated by γ rays and exposed to a combination of γ irradiation followed by thermal aging at 90℃. Swelling pressure tests were then conducted. Results showed that irradiation and thermal aging reduced the swelling pressure of GMZ bentonite. Notably, after irradiating and storing at (25 ± 3)℃ for 360 d, the swelling pressure increased by 51.59 % compared with that of the γ-irradiated sample; the swelling pressure decreased by 75.9 % for the γ irradiation–thermal sequentially aged sample (heated for 180 d). Partial chemical-bond destruction, the transformation of structure trivalent to divalent iron, and the water redistribution in the bentonite were the primary drivers of the decrease in the swelling pressure of GMZ bentonite.

引用次数: 0

Assessment and predicting the axial power distribution effect on the thermal-mechanical parameters of the NuScale nuclear reactor core loaded with TVS-2 M fuel assemblies as well as axial Offset optimizing for load-following operation

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-26 DOI: 10.1016/j.nucengdes.2025.114009

M.H. Zahedi yeganeh, G.R. Ansarifar, H.Zayermohammadi Rishehri

This study evaluates and examines the thermal–mechanical behavior of a NuScale reactor core which utilizes TVS-2 M hexagonal fuel assemblies. The efficiency of the fuel rods is validated using the FRAPCON code. Initially, the reactor’s core is modeled with the MCNP code to locate the control banks. The design phase ensures the capability to shut down the reactor in two scenarios. In the Hot Zero Power (HZP) scenario, MCNP simulation reveals a sub-critical state with a multiplication factor of 0.94481 ± 0.00023. In the Cold Zero Power (CZP) scenario, the multiplication factor of 0.9935 ± 0.00023 confirms the adequacy of control assemblies. Subsequently, a thermal–mechanical analysis is conducted on the fuel rod over 1330 days, confirming its acceptable design and operational effectiveness in the core. Also, one of the parameters that can be examined during reactor control and load-following operations is Axial Offset (AO). Therefore, the study investigates the impact of AO on fuel rod’s thermal–mechanical changes. The MCNP code was used to simulate control rod inputs and obtain power distribution data for each AO deviation. Based on assessments regarding the association between AO and the thermal–mechanical characteristics of fuel, it has been determined that the impact of power distribution increases significantly over time, particularly towards the end of the operational period. Afterward, based on FRAPCON results, an artificial neural network (ANN) estimator is developed to predict thermal–mechanical parameters at the beginning of the cycle (BOC). The ANN proves to be a powerful method for estimation. By employing the ANN estimator and exploring different cost functions based on thermal–mechanical parameters, the optimal AO is determined using a genetic algorithm, which enhances the reactor’s performance, particularly in load-following operations. The attained optimal AO value for various cost functions are as follows: −0.10316, −0.19635, and −0.25817. This approach allows for the selection of the most efficient AO, leading to improved performance of the NuScale reactor core loaded with TVS-2 M hexagonal fuel assemblies. Indeed, optimization of AO is very important and useful for load-following operation.

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引用次数: 0

Analyses of the MELCOR capability to simulate integral PWR using passive systems in a DBA scenario

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-24 DOI: 10.1016/j.nucengdes.2025.114004

M. Principato , F. Giannetti , M. Imperatori , M. D’Onorio , M. Garcia , L.E. Herranz , A. Bersano , F. Mascari

This paper is focused on a thermal–hydraulic transient analysis of a generic 300 MWe integral Pressurized Water Reactor (iPWR) based on a passive safety mitigation strategy and dry containment. The main target of the paper is to analyze the MELCOR code’s capability to simulate the operation of the passive systems in Small Modular Reactor (SMR) configuration and the consequent plant behavior. The nodalization and the modeling approach used in the MELCOR code have been described in the present work, and a double-ended rupture of the Direct Vessel Injection (DVI) line has been postulated as initiating event.

The results of the analysis demonstrated that the MELCOR code can qualitatively replicate the dominant phenomena that drive the passive mitigation strategy’s operation. Additionally, the core reflooding was made possible by the operation of the available safety systems, which was adequate to prevent severe accident conditions during the entire simulated transient. The activity has been developed in the framework of SASPAM-SA Horizon Euratom project as the base for further derivative activities focused on assessing the capability of the MELCOR code to simulate the phenomena taking place in postulated plausible severe accident scenarios in iPWR.

{"title":"Analyses of the MELCOR capability to simulate integral PWR using passive systems in a DBA scenario","authors":"M. Principato , F. Giannetti , M. Imperatori , M. D’Onorio , M. Garcia , L.E. Herranz , A. Bersano , F. Mascari","doi":"10.1016/j.nucengdes.2025.114004","DOIUrl":"10.1016/j.nucengdes.2025.114004","url":null,"abstract":"<div><div>This paper is focused on a thermal–hydraulic transient analysis of a generic 300 MWe integral Pressurized Water Reactor (iPWR) based on a passive safety mitigation strategy and dry containment. The main target of the paper is to analyze the MELCOR code’s capability to simulate the operation of the passive systems in Small Modular Reactor (SMR) configuration and the consequent plant behavior. The nodalization and the modeling approach used in the MELCOR code have been described in the present work, and a double-ended rupture of the Direct Vessel Injection (DVI) line has been postulated as initiating event.</div><div>The results of the analysis demonstrated that the MELCOR code can qualitatively replicate the dominant phenomena that drive the passive mitigation strategy’s operation. Additionally, the core reflooding was made possible by the operation of the available safety systems, which was adequate to prevent severe accident conditions during the entire simulated transient. The activity has been developed in the framework of SASPAM-SA Horizon Euratom project as the base for further derivative activities focused on assessing the capability of the MELCOR code to simulate the phenomena taking place in postulated plausible severe accident scenarios in iPWR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114004"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682450","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}

引用次数: 0

Justification of RCCAs lifetime extension at operating Ukrainian NPPs. Summary calculations

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-24 DOI: 10.1016/j.nucengdes.2025.114008

Valeriy Zuyok , Oleksandr Mazurok , Oleg Godun , Mykola Chaikovskyi , Anton Makarenko , Vadym Ivanov , Valodymyr Zigunov , Mykhaylo Tretyakov

One of the ways to temporarily meet the needs of NPPs for rod cluster control assemblies (RCCAs) is to extend their lifetime. The justification for extending the lifetime of RCCAs should be based on the justification of mechanical reliability of control rods (CRs) and RCCAs in general, fulfillment of the criteria for physical efficiency of RCCAs, thermotechnical characteristics of RCCAs, taking into account changes in materials characteristics during the entire previous period of operation in the reactor core.

In the course of the research, the operational data (RCCAs positions and power unit capacity) for each of the 61 RCCAs of all 13 VVER-1000 power units for the last almost 10 years were analyzed and systematized. The maximum fast neutron flux and ¹⁰B burnup were calculated for each RCCA, and the margin before reaching the limit values was also calculated.

The results of the performed research enabled to extend the lifetime of almost all RCCAs in VVER-1000 reactors of Ukrainian NPPs from 25,500 to 38,000 h in the automatic control group and from 75,600 to 113,500 h in the reactor scram group, except for twelve RCCAs from RNPP Unit 4 and six RCCAs from SUNPP Unit 1. An individual (separate) assessment of the lifetime was performed for these RCCAs, throughout which the operational integrity criteria for the maximum value of fast neutron flux in the lower part of CR cladding and ¹⁰B burnup will be met.

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引用次数: 0

Effect of spray flow rate on pressure and temperature distribution in SMR containment

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-23 DOI: 10.1016/j.nucengdes.2025.114005

Jinglin Cao, Xuefeng Lyu, Fenglei Niu, Jialei Chen

In the case of a direct vessel injection (DVI) line break in small modular reactors, the key to ensuring reactor safety is a timely and effective suppression of the pressure and temperature rise in the containment. In this paper, GASFLOW was utilized to analyze the influence of the internal spray mass flow rate on the pressure suppression and to compare 3D temperature distribution in the containment following the double-ended DVI line rupture. Results showed that the spray system significantly reduced the pressure and temperature in the containment in the initial phases. As the accident progressed, the impact of varying spray flow rates on the containment pressure and temperature gradually diminished, temperature distribution became more uniform, and the condensation effect of the spray ultimately stabilized. These findings substantiate the efficacy of the spray system and reveal a positive correlation between spray flow rates and more evident pressure suppression and cooling effects.

引用次数: 0

Uncertainty-aware prediction of Peak Cladding Temperature during extended station blackout using Transformer-based machine learning

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-22 DOI: 10.1016/j.nucengdes.2025.113984

Tran C.H. Nguyen , Aya Diab

Accurate prediction of the Peak Clad Temperature (PCT) may be used to evaluate the efficacy of operator mitigation actions during extended Station Blackout (SBO) scenarios. In this study, we propose a two-stage machine learning (ML) framework that integrates classification and regression to forecast PCT. While the classification stage identifies whether mitigation efforts succeed or fail, the regression stage provides precise multi-step PCT predictions. Our framework leverages advanced ML models, including Transformer architectures, Attention mechanism, and Long Short-Term Memory (LSTM) networks, alongside the Best Estimate Plus Uncertainty (BEPU) approach. To account for the underlying uncertainty and generate confidence intervals, we incorporate Monte Carlo (MC) Dropout. By integrating BEPU with machine learning and uncertainty quantification, our model produces reliable temperature forecasts despite the system’s inherent complexity and nonlinearity with R² values exceeding 0.98 for 60-, 120-, and 240-step time frames. Notably, the LSTM-Transformer model proves to be the most effective, even for longer prediction horizons. The developed framework serves as a powerful real-time decision support tool for operators, for accurate prediction and effective mitigation of critical conditions like extended SBO events.

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