Release Source Term Research Articles

Source term inversion of nuclear accidents based on ISAO-SAELM model

The release source term of radioactivity becomes a critical foundation for emergency response and accident consequence assessment after a nuclear accident Rapidly and accurately inverting the source term remains an urgent scientific challenge. Today source term inversion based on meteorological data and gamma dose rate measurements is a common method. But gamma dose rate actually includes all nuclides information, and the composition of radioactive nuclides is generally uncertain. This paper introduces a novel nuclear accident source term inversion model, which is Improve Snow Ablation Optimizer-Sensitivity Analysis Pruning Extreme Learning Machine (ISAO-SAELM) model. The model inverts the release rates of 11 radioactive nuclides (I-131, Xe-133, Cs-137, Kr-88, Sr-91, Te-132, Mo-99, Ba-140, La-140, Ce-144, Sb-129). It does not require the use of the physical field of the reactor to obtain prior information and establish a dispersion model. And the robustness is validated through noise analysis test. The mean absolute errors of the release rates of 11 nuclides are 15.52 %, 15.28 %, 15.70 %, 14.99 %, 14.85 %, 15.61 %, 15.96 %, 15.42 %, 15.84 %, 15.13 %, 17.72 %, which show the significant superiority of ISAO-SAELM. ISAO-SAELM model not only achieves notable advancements in accuracy but also receives validation in terms of practicality and feasibility.

Evaluation of Thermoacoustic Instability for Chemically Reacting Flows Using Large-Eddy Simulations

Abstract Thermoacoustic instability for chemically reacting flows was investigated using large eddy simulations coupled with a lookup table for turbulence-chemistry closure. The onset of instability was evaluated from pressure fluctuations, as well as standard and extended Rayleigh criterion, as suggested in literature. Two configurations were considered, namely, a canonical Rijke tube and a simplified can combustor with a swirling flow injector representing a complex generalized geometry. For the Rijke tube, premixed and nonpremixed combustion models were applied for identical fuel flowrate, resulting in different thermoacoustic outcomes due to differences in reaction rates of the two flame regimes. Results from the Rijke tube case agree with analytic thermoacoustic theory. For the can combustor, only premixed chemistry was considered as it better represents the experimental conditions, and the first resonant pressure mode aligns reasonably with published experimental data. Findings suggest that, if thermoacoustic instability is detected, the resonant frequency can be deduced from the fluctuations of the pressure, heat release, or acoustic source term. However, even though the resonant frequency is correctly identified, fluctuation data alone is insufficient to identify the onset of thermoacoustic instability, requiring the additional application of Rayleigh criterion. Finally, this study concludes that, for the range of configurations evaluated here, the standard Rayleigh criterion is sufficient to determine the onset of thermoacoustic instability, so the extended Rayleigh criterion is not always necessary, in contrast to suggestions from previous work. This conclusion is significant because the standard Rayleigh criterion is the only practical evaluation for physical experiments.

Quantitative visualization of a radioactive plume with harmonizing gamma-ray imaging spectrometry and real-time atmospheric dispersion simulation based on 3D wind observation

ABSTRACT To propose a novel monitoring method for the quantitative visualization of 3D distribution of a radioactive plume in the vicinity of release point and source term estimation of released radionuclides, the feasibility of its analysis method is demonstrated by preliminary test. The proposed method is the combination of gamma-ray imaging spectroscopy with the electron tracking Compton camera (ETCC) and real-time high-resolution atmospheric dispersion simulation based on 3D wind observation with Doppler lidar. ETCC can acquire the angle distribution images of direct gamma-ray from a specific radionuclide in a target radioactive plume by quantitatively capturing gamma-ray incidence angles and spectrum distributions. The 3D distribution of each radionuclide in the radioactive plume is inversely reconstructed from direct gamma-ray images of each radionuclide by several ETCCs located around the target by harmonizing with the air concentration distribution pattern of the plume predicted by real-time atmospheric dispersion simulation. A prototype of the analysis method was developed, showing a sufficient performance in several test cases using hypothetical data generated by numerical simulations of atmospheric dispersion and radiation transport. The proposed method is expected to become an innovative monitoring framework for nuclear emergency.

Study on the influence of source term release on accident consequences in severe accidents for small pressurized water reactor

In this paper, a small PWR with no pressure vessel external cooling system (ex-vessels) is used as the research object. It is found that for the small low burnup marine reactor without ex-vessels, the more core releases at an early stage, the later core damage occurs. Considering in the case of large break loss of coolant and loss of power accident, and the ORIGEN program is coupled with the MELCOR program is used as a severe accident process simulation tool. The significant parameters have been took into account, include core melting time, core damage rate, and lower head failure time, etc. It could be found through calculation that, for a certain reactor, in case of the reactor core meltdown happened, different released amount of source term will results in a difference of core heat release in the pressure vessel. For the small reactors without ex-vessels, the amount difference of source term release can reach 20%, and the lower head failure time can be delayed by 70%.

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Consequence assessment considerations for toxic natural gas dispersion modeling

The numerical simulation of gas dispersion and estimation of consequence impact is of importance in Oil and Gas industry's process safety management. For natural gas fields with toxic components like Hydrogen sulfide, the toxicity impact zone drives business decisions related to equipment design, facility siting, layout, land use planning, and emergency response measures. Proprietary tools or empirical models which are calibrated using experiment database are often used for carrying out consequence modeling.The selection of a software tool and a suitable dispersion model, based on the cloud behavior, at the source of dispersion is critical for the impact zone estimation. It is observed that, the fluid phase and the cloud density are key for determining the appropriate dispersion model. Incorrect parameter selection could lead to an inaccurate consequence impact zone estimation. This in turn could result in disproportionate process risk management efforts especially for toxic impacts from exposure to a very low concentration.This paper discusses the results from consequence modeling studies done for a selected set of toxic natural gas release events related to onshore pipeline transfer using approved software. The study analyses the modeling inputs, parameters and determines the key release source terms and atmospheric parameters that impacts the estimation of impact zones. The study determined that the natural gas dispersion behaviour is dependent on the natural gas molar mass and the composition of Hydrogen sulfide. The study provides the guidance on overcoming uncertainty in dispersion modeling through sensitivity assessments and lists key parameters to be subjected for toxic natural gas dispersion modeling sensitivity analysis.

Analysis of an unmitigated 2-inch cold leg LOCA transient with ASTEC and MELCOR codes

The analyses of postulated severe accident sequences play a key role for the international nuclear technical scientific community for the study of the effect of possible actions to prevent significant core degradation and mitigate source term release. To simulate the complexity of phenomena involved in a severe accident, computational tools, known as severe accident codes, have been developed in the last decades. In the framework of NUGENIA TA-2 ASCOM project, the analysis of an unmitigated 2-inch cold leg LOCA transient, occurring in a generic western three-loops PWR-900 MWe, has been carried out with the aim to give some insights on the modelling capabilities of these tools and to characterize the differences in the calculations results. The ASTEC V2.2b code (study carried out with ASTEC V2, IRSN all rights reserved, [2021]), and MELCOR 2.2 code have been used in this code-to-code benchmark exercise. In the postulated transient, the unavailability of all active injection coolant systems has been considered and only the injection of accumulators has been assumed as accident mitigation strategy.

Realistic evaluation of source terms for steam generator tube rupture accidents

In the existing Probabilistic Safety Assessment (PSA) model, all the Steam Generator Tube Rupture (SGTR) sequences have been assumed to have identical fission product release characteristics without considering the sequences of no source term release resulting in a conservative evaluation. Since the Korean Nuclear Safety Act was revised in 2016, a realistic method that can reasonably evaluate the risk became needed to satisfy the PSA targets. Therefore, a new method is proposed to improve the Level 2 PSA model (decomposition event tree, containment event tree, source term category logic diagram) that can identify the SGTR sequences without fission product release to the environment. The sequences are classified by dividing the Source Term Category (STC) of SGTRs into SGTR_CON (containment) and SGTR_ENV (environment). If fission product is trapped inside the containment building and not released to the environment, it is classified as SGTR_CON. Otherwise it is classified as SGTR_ENV. About 72% of the existing STC for SGTR is treated as SGTR_CON, which does not release fission product to the environment. If SGTR STCs are divided into SGTR-CON and SGTR-ENV, then a more reasonable source term evaluation of SGTR sequences is expected.

Impact of source term release characteristics of nuclear plant on the performance of EURDEP to identify radioactive plumes

This study aims at analysing the performance of the European Radiological Data Exchange Platform (EURDEP) at identifying, delineating, and tracking different radioactivity clouds moving across a territory. To this purpose, the atmospheric dispersion of 21 hypothetical STs, calculated by the Modular Accident Analysis Program (MAAP) code, has been simulated under the same meteorological scenario by the JAVA-based RODOS decision support system (JRODOS). EURDEP's performance is evaluated using spatial and temporal parameters. The present analysis shows a large impact of the total amount of activity and the timing of the release peak on EURDEP's performance. Monitoring stations are able to detect quite rapidly the release (first alarm), and presents more variation in the time needed to detect the peak, and to delineate and track the plume. A two-dimensional analysis is suggested to illustrate strengths and weaknesses of EURDEP to face different types of STs. These results highlight the need to link the understanding of ST characteristics with meteorological conditions to limit the environmental and health effects associated with radioactive emissions into the atmosphere. The present study aims at presenting a methodology to analyse the performance of EURDEP in other territories.

Study on Release and Migration of Radionuclides under the Small Break Loss of Coolant Accident in a Marine Reactor

Using severe accident analysis program MELCOR, the small break loss of coolant accident (SBLOCA) analysis model was established for a marine reactor. The release and migration of radionuclides were analyzed during a severe accident induced by SBLOCA. The analysis of the hydrogen source term release showed that the maximum hydrogen release amount was 248.567 kg, and the hydrogen release amount accounted for less than 4% of the air volume. So, there would be no danger of hydrogen explosion accidents. The research mainly focused on the behaviors of the release, the transport, the retention, and the final distribution of inert gases represented by Xe, volatile gases represented by CsI, and nonvolatile nuclides represented by Ba. The results showed that the reactor core exposed completely with a lagging by 510 s and the initial release time of nuclides was lagged by 1916 s. The release shares of Xe in the primary circuit system, the containment, and the environment were 0.013%, 0.06%, and 32.71%, respectively. Also, Ba shared 0.016%, 0.0032%, and 3.28%, respectively, and CsI shared 0.0145%, 0.0012%, and 2.845%, respectively.

Preliminary study on inversion model of radioactive release source terms

Objective Based on the monitoring data, Gaussian plume model and optimization algorithm, the inversion model of radioactive release source term was preliminarily studied. Methods By comparing the application conditions and performance of several optimization algorithms, the accuracy of source term inversion results of different algorithms under the same conditions was compared and verified. Results The preliminary study of radioactive release source term inversion model shows that it is feasible to use the least square method to optimize the inversion model based on Gaussian plume. The accuracy of the optimized model is 0.92 times of the measured value for flat terrain, and 1.5 times of the measured value for complex terrain, whose error is relatively large because of the limitation of the model itself. Conclusion The results show that the optimized source term inversion model based on Gaussian plume and least square method can better meet the needs of nuclear accident consequence assessment and emergency decision-making in terms of computational performance and accuracy. 摘要： 目的 基于监测数据、高斯烟羽模型和优化算法, 初步研宄放射性释放源项反演模型。 方法 通过比较几种优 化算法的适用条件和性能, 并对相同条件下不同算法源项反演计算结果的准确性进行比较验证。 结果 通过对放射 性释放源项反演模型初步研宄, 发现基于高斯烟羽的反演模型, 采用最小二乘法优化后进行释放源项反演计算是可行 的。该优化模型对于简单地形下的模拟结果精度较高, 为实测值的 0.92 倍; 对于复杂地形条件下的模拟结果受模型本 身限制误差相对较大, 为实测值 1.5 倍。 结论 研宄表明高斯烟羽结合最小二乘法的优化源项反演模型在计算性能和 精度上可更好地满足核事故后果评价和应急决策需求。

Application of Modified Genetic Algorithm in Source Term Inversion of Nuclear Accident on Offshore Platform

Wang, X.; Wang, B.; Wang, Y.; Yang, Y.; Du, F., and Han, F., 2020. Application of modified genetic algorithm in source term inversion of nuclear accident on offshore platform. In: Yang, Y.; Mi, C.; Zhao, L., and Lam, S.(eds.), Global Topics and New Trends in Coastal Research: Port, Coastal and Ocean Engineering. Journal of Coastal Research, Special Issue No. 103, pp. 812–816. Coconut Creek (Florida), ISSN 0749-0208.In this paper a source term inversion model is built to study the source release during severe accident in nuclear power plant. The model is based on the combination of genetic algorithm and Nelder-Mead simplex algorithm. It uses the off-site monitor data and the meteorological data to estimate the amount of radiological materials released into environment during severe accident. This method can support the rapid obtain of source term during nuclear emergency and become a supplement to the core damage evaluation method. Based on numerical analysis experiments, indoor wind tunnel experiments and field tracing experiments, the established inversion model is verified. The verification results show that the source term inversion model established in this paper can quickly and accurately give the release source term and the max error belows 48%. This model can be used in the consequences evaluation of nuclear accidents, provide suggestions to accident emergency decision-making.

SOURCE TERM ASSESSMENT FOR 100 MWE PRESSURIZED WATER REACTOR

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

Numerical investigation on combustion instability modeling in a lean premixed gas turbine combustor combining finite element analysis with local flame transfer function

The present work suggests a numerical approach using an FEM (finite element method) code for the eigenvalue calculation of the thermoacoustic combustion instabilities coupling with the local flame transfer function in a generic lean premixed combustor. The URANS (unsteady Reynolds average Navier-Stokes) simulation shows a good agreement in capturing steady and unsteady flame geometries with the measurement in the literature. Also, global and local parameters of flame responses to acoustic velocity fluctuations undergo quantitatively similar changes in prediction and in the experiment. The numerically determined flame transfer function and gas properties are incorporated in an in-house Helmholtz solver to compute eigenvalues of the target combustor. In the solver, the Helmholtz equation is discretized with a Galerkin FEM on a multi-dimensional hybrid unstructured mesh. The nonlinearity related with the heat release source term is treated with an iterative methodology, and the large scale eigenvalue calculation is performed using the shift-invert approach available in the ARPACK (ARnoldi PACKage) library. The predictive capabilities of the current modeling approach are validated against the measured mode frequency and instability growth rate. In addition, the prediction results show the effect of spatially-distributed heat release on the thermoacoustic model accuracy.

Estimation of Tritium and Dust Source Term in European DEMOnstration Fusion Reactor During Accident Scenarios

The safety features of the future nuclear fusion reactors are one of the key issues for their attractiveness if compared with the fission plants. In fusion devices, accidents with high release of radioactive materials have low probabilities because the most part of abnormal transients lead to passive plasma shutdown. It does not mean that radiological source terms such tritium and activated dust are not generated and released, but their inventory does not increase during abnormal events. Therefore, the source term inventory has to be assessed during normal operation and traced when accidents occur. For this reason, a study for qualification and quantification of the tritium and dust source term (DTS) was established with the aim to understand their production, deposition, and penetration in the vacuum vessel (VV) and in the breeding blanket (BB). The main concern is source term release during the main accident scenarios to comply with a future licensing process. In case of abnormal event scenarios, the source term inventory involved in the release changes and requires a different confinement approach and mitigation. For the estimation of the source term in the DEMOnstration Fusion Power Station (DEMO), a methodology was developed. The methodology scales the tritium and DTS inside the VV from the International Thermonuclear Experimental Reactor, the European Power Plant Conceptual Study, and reports the tritium generated inside the breeder blanket from data quantified in other studies for DEMO. In this article, the methodology was updated and tritium and DTS for DEMO 2016 design were estimated. Moreover, the tritium and dust release pathways were highlighted according to different accidental scenarios. These results were obtained for all blanket concepts, which are analyzing in the ongoing DEMO EUROFusion project. The values estimated in this article will be used in the safety analyses to evaluate releases or to quantify the operational limits starting from values postulated in International Thermonuclear Experimental Reactor.

A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor

This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.

Contribution of production and loss terms of fission products on in-containment activity under severe accident condition for VVER-1000

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

Анализ тяжелой запроектной аварии ректорной установки ЭГП-6 билибинской АЭС. Формирование источника выброса

Анализ тяжелой запроектной аварии ректорной установки ЭГП-6 билибинской АЭС. Формирование источника выброса

Development of a Procedure for Evaluating the Impacts of the Accidental Emission of Hazardous Chemicals, Case Study in Ho Chi Minh City, Vietnam.

The study aims at developing a process to evaluate the impacts of the accidental emission of hazardous chemicals. The proposed process consists of four basic steps: i) identifying risks/hazards; ii) development of the worst-case scenario; iii) simulating the emission and dispersion of the toxic chemicals; and iv) assessing the severity of the impact to the people and the surroundings. It makes use of different techniques including accidental release source term, atmospheric dispersion modeling and results in the concentration and extent of the toxic chemicals in the atmosphere for either the direct evaporation of toxic chemicals as a primary emission or the dispersion of toxic chemicals as a domino effect of a fire or explosion accident. This process has been applied in a contrived case study in Ho Chi Minh City, Vietnam. In a suppositious accident of p-xylene spill from a pesticide factory, the assessment for the worst-case scenario showed that p-xylene concentration in the atmosphere could reach up to 8,882,381 µg/m3, that is higher than Protective Action Criteria for Chemicals-level 2 but far lower than the level 3. p-Xylene from the accident could disperse more than 20 km from the site, to a highly populated area with a large number of sensitive social economic object. The results of this assessment provide helpful information for the development of accidental response plan in the practical cases or supports the training for accident prevention and responses.

Recirculation of biomass ashes onto forest soils: ash composition, mineralogy and leaching properties

In Denmark, increasing amounts of wood ashes are generated from biomass combustion for energy production. The utilisation of ashes on top of forest soil for liming purposes has been proposed asan alternative to landfilling. Danish wood ash samples were collected and characterised with respect to chemical composition, mineralogy and leaching properties (batch leaching at L/S 2 and 10L/kg, and pH-dependent leaching at 10L/kg). Large variations in the ash liming properties were observed (ANC7.5: 1.8–6.4meqH+/g), indicating that similar soil application dosages may result in different liming effects. High contents of Ca, Si, P, K and Mg were observed in all samples, while the highest contents of S and N were found in fly ashes and mixed ashes (combination of fly and bottom ashes). Similarly, the highest contents of some trace metals, e.g. Cd, Mo and Se, were observed for fly ash. Releases of major, minor and trace elements were affected significantly by pH: high releases of PO43−, Mg, Zn, Cu and Cd were found for acidic conditions relevant to forest soils, while the highest releases of Mo and Cr were observed in alkaline conditions. Mineral phases were selected based on XRD analyses and the existing literature, and they were applied as inputs for the geochemical modelling of pH-dependent leaching. Mineral dissolution was found adequate for a wide range of major elements and nutrients, while the description of trace elements could be done only for parts of the pH-range. Content and leaching of PAHs were observed below detection limits. The source-term release of Ca, K, Mg, Mn, and P in acidic conditions relevant to forest soils was higher than ten years of atmospheric deposition, in contrast to the relatively low release of Al, Fe and Na. The potential release of Cd was found to be the most critical element compared with soil quality criteria, whereas the maximum theoretical loads of Ba, Cd, Cr, Sr, Mo, Ni, Pb, Sb, Se, Sn and V were relatively low.