اميدصفرزادهOmidSafarzadehمهندسی پزشکیدانشياردانشکده فني و مهندسيhttps://scholar.google.com/citations?user=mjCoZ6IAAAAJ&hl=enhttp://ref.shahed.ac.ir/safarzadehhttp://research.shahed.ac.ir/WSR/WebPages/Teacher/TFa.aspx?TID=348http://research.shahed.ac.ir/WSR/WebPages/Report/ResearchActivity.aspx?TeacherID=34831490Physical probability table applied to response matrix-collision probability method for direct whole-core transport calculationsPhysical probability table applied to response matrix-collision probability method for direct whole-core transport calculations1394-12-112016-03-01In this study, we applied physical probability table to response matrix-collision probability method to treat and evaluate self-shielded resonance cross-sections for whole-core transport calculations. To attain accurate results, especially in resonance treatment, the conservation laws of probabilities are derived and satisfied using HELIOS normalization scheme. The subgroup weights are efficiently evaluated using the conjugate gradient method based on the normal equations (CGNR). The resonance interference effects are considered. We used Segev method to interpolate the resonance integrals (RIs), in order to accurately obtain the subgroup weights. Finally, a validation is presented where the effective absorption and production cross-sections and the infinite multiplication factors are compared with exact values obtained using NJOY and the equivalence theory for homogeneous media. Other results, such as the Rowland’s benchmark for three assembly production cases, are also presented and compared with those produced by Monte Carlo method. The results indicate the developed method can accurately represent the selfshielding effects.O. SafarzadehArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=31490http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=314900306-45491873-2100http://www.sciencedirect.com/science/article/pii/S030645491500557531907Application of h -adaptivity to the interface current collision probability methodApplication of h -adaptivity to the interface current collision probability method1395-02-172016-05-06In this paper, we presented an adaptive algorithm for the interface current collision probability method. The algorithm is based on a posteriori estimation scheme to obtain the residual errors in the problem of interest. The discretization error is obtained by a flux gradient approach based on the net current leakage of nodes, and also by the local balance formulated through the residual of the particle conservation equation. To implement the adaptive approach, a computer program has been developed to solve the Boltzmann transport equation (BTE). The suitability of the adaptive refinement strategy is demonstrated by performing a series of numerical experiments in a simple shielding problem and reactor core geometry that exhibit strong heterogeneity and flux gradients. The effectiveness of the method is assessed by comparing the number of meshes required with that necessary using uniform refinement to indicate what savings can be achieved by adaptive refinement.O. SafarzadehArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=31907http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=319070306-45491873-2100http://www.sciencedirect.com/science/article/pii/S030645491630221385409تعيين محدوده مجاز تغييرات توان نسبي بافرض حادثه انسداد در راکتور VVER-1تعيين محدوده مجاز تغييرات توان نسبي بافرض حادثه انسداد در راکتور VVER-11396-11-082018-01-28همانطور که میدانیم، در مطالعه و ارزیابی عملکرد ایمن راکتورهای هستهای، ارزیابی حوادث یکی از زمینههای بسیار حائز اهمیت است.پدید امدن نقص در سیستم انتقال حرارت (نظیر ازدست رفتن خنک کننده LOCAو یا جریان خنک کننده )LOFAگروهی از حوادث ممکن در راکتورهای هسته ای را تشکیل میدهند. از جمله حالات گذرای ناشی از نقص در سیستم خنک کننده، حادثه انسداد موضعی مسیر جریان خنک کننده است که میتواند باعث کاهش میزان جریان خنک کننده در یک یا تعدادی از مجتمعهای سوخت راکتور شود. حادثه انسداد میتواند به دلائل مختلفی رخ دهد. یکی از مهمترین این عوامل، شکست و فرو افتادن قطعاتی از اجزایی از مدار اول در مدار اول راکتور است. در این مطالعه حادثه انسداد موضعی مسیر جریان در ورودی یک مجتمع سوخت داغ در ضرایب مختلف توزیع توان نسبی مجتمع سوخت با استفاده از کد COBRA-ENو برنامهای که با نرم افزار متلب برای بررسی انسداد یک مجتمع سوخت داغ تدوین شده است،بررسی شده است. ابتدا به منظور ارزیابی کد COBRA-ENو برنامه تدوین شده، کارکرد شرایط عادی قلب راکتور VVER-1000مدل شده و نتایج با FSARمقایسه شده است تا صحت مدلسازی توسط کد COBRA-ENو برنامه مذکور تایید گردد. در ادامه در ضرایب توزیع نسبی توان مختلف و مقدار انسدادهای مختلف مدلسازی انسداد مجتمع سوخت انجام شده است. نتایج نشان داده است که همراه با در نظر گرفتن سایر عوامل موثر در تولید و توزیع توان در قلب نظیر توزیع شار نوترونی و ضریب تکثیر ( ،) Keffمجتمعهای سوخت داخل قلب راکتور باید طوری چیده شوند تا مقدار ضریب توزیع توان نسبی مجتمع سوخت داغ همواره بین دو مقدار 0/71و 1/85باشد. دراینصورت وقوع حادثه انسداد در مجتمع سوخت داغ اثر خاصی در عمکرد راکتور ندارد.rahman gharari, Omid Safarzadeh, Naimaddin Motaji KajuriArticleتابش و فناوري هسته ايژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=85409http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=854092423-6661____-____http://jrnt.guilan.ac.ir/article_2971.html85532The ?–synthesis and analysis of water level control in steam generatorsThe ?–synthesis and analysis of water level control in steam generators1397-11-232019-02-12The robust controller synthesis and analysis of the water level process in the U-tube system generator (UTSG) is addressed in this paper. The parameter uncertainties of the steam generator (SG) are modeled as multiplicative perturbations which are normalized by designing suitable weighting functions. The relative errors of the nominal SG model with respect to the other operating power level models are employed to specify the weighting functions for normalizing the plant uncertainties. Then, a robust controller is designed based on m-synthesis and D-K iteration, and its stability robustness is verified over the whole range of power operations. A gain-scheduled controller with H∞ -synthesis is also designed to compare its robustness with the proposed controller. The stability analysis is accomplished and compared with the previous QFT design. The meanalysis of the system shows that the proposed controller has a favorable stability robustness for the whole range of operating power conditions. The proposed controller response is simulated against the power level deviation in start-up and shutdown stages and compared with the other concerning controllers.Omid Safarzadeh, Mohammad Hosein Kazemi, Ahmed SalehiArticle Nuclear Engineering and Technologyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=85532http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=855321738-57331738-5733http://dx.doi.org/10.1016/j.net.2018.09.01895848Fractional order PID control of steam generator water level for nuclear steam supply systemsFractional order PID control of steam generator water level for nuclear steam supply systems1397-11-122019-02-01The main part of nuclear powered system is the nuclear steam supply system (NSSS) to produce the steam needed for electricity or cogeneration. The NSSS mainly contains the fission reactor, based on pressurized water reactors or small modular reactors, and the steam generator. One of the most important control strategies in the NSSS is water level regulation to keep the water level of the steam generator around programmed setpoint because violation of the level jeopardizes the plant availability in making the electricity and supplying the heat for industrial applications. The appearance of fractional calculus made it possible to realize the control design requirements in optimum and efficient ways. In this paper, we proposed a gain scheduled Fractional order PID (FOPID) control system for steam generator level control system in entire operating range. Nelder-Mead tuning strategy which takes into account the desired gain- and phase- margin with integral time absolute error (ITAE) performance index has been devoted to adjust the FOPID parameters. Simulation results indicate that the new control system can enhance the transient response in contrast to the conventional PID. The performance comparison of the FOPID regulator with the classical PID shows that the fractional controller has huge superiority which is comparable with the advanced quantitative feedback theory (QFT) controller. The Nyquist stability analysis is also provided to demonstrate the favorable robust stability of the gain scheduled FOPID controller over the wide range of operating conditions.Ahmad Salehi, Omid Safarzadeh, Mohammad Hossein KazemiArticleNuclear Engineering and Designژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=95848http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=958480029-5493____-____http://dx.doi.org/10.1016/j.nucengdes.2018.11.040106434Accurate full superfine group resonance self-shielding method with fission source and upscattering effects in thermal latticeAccurate full superfine group resonance self-shielding method with fission source and upscattering effects in thermal lattice1398-01-192019-04-08This paper focused upon an approach to the promotion of the resonance self-shielding treatment to overwhelm the reaction rates deviation that mitigating frequent assumption of the ultrafine energy group method. Our first major finding is implementation of both the dependency of the fission source on incident neutron energy and thermal upscattering matrix at desired energy mesh on effective cross section. Moreover, the escape cross section correction (EC) method reflecting geometrical effect of 2D lattice is proposed to revise 1D cylindrical model on coarse group cross section collapsing by SPH method. After verifying the superfine library generation module, called UFEMR, various benchmark calculations for pin cell and fuel assembly are tested against the results of corresponding Monte Carlo solutions. The average relative error of the reaction rates by developed cross section generation module and use of the proposed superfine group structure is around 0.74. The regionwise flux and reaction rates are in good agreement with Monte Carlo reference solutions.O. SafarzadehArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=106434http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1064340306-45491873-2100http://dx.doi.org/10.1016/j.anucene.2019.04.002116542The surface current method in gray Dancoff factor calculationThe surface current method in gray Dancoff factor calculation1398-02-252019-05-15The Dancoff correction is important in the calculation of the effective cross section of resonant isotopes in a heterogeneous system. Although the neutron currentmethod is a simple and straightforward approach to estimate the Dancoff factor, its use is limited to the black Dancoff factor. In this paper, we expand the current method used to determine both the black and gray Dancoff factors. The method developed also relies on a neutron transport solver, where a fixed source on a fuel rod surface has an outward direction, a cosine distribution, and a constant shape. The detector is located on the surface of the rods to measure incoming and outgoing currents; therefore, there is no need to calculate the chord length, and the development, validation, and verification of the code canbe omitted. The mathematical foundation of the suggested method is derived using the integral transport equation. The effects of the moderator and lattice configuration are followed by a sensitivity analysis of the Dancoff factor for several problems, including pressurized water reactor and cluster fuel assemblies. The maximum and average relative errors of the calculated results are approximately 0.3 and 0.05, respectively.Omid SafarzadehArticleNuclear Science and Techniquesژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=116542http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1165421001-80422210-3147http://dx.doi.org/10.1007/s41365-019-0622-2137239Heterogeneous reactor core transport technique using response matrix and collision probability methodsHeterogeneous reactor core transport technique using response matrix and collision probability methods1392-09-142013-12-05In this paper a whole-core transport technique using response matrix and collision probability (CP) methods is presented for large-scale, highly-heterogeneous reactors. Integral transport method has been used to provide sufficient accuracy for response matrix formation of pin cell sized node with considerably less computational expense. Ray tracing is efficiently applied using macro-bands. For practical application, double P2 (DP2) Legendre polynomial expansion is applied to approximate interface angular flux that is used to couple nodes. The proposed method is based on a sound mathematical foundation and leads to dramatically reduced memory requirements in contrast to the conventional transport method. This method is also applied to several problems such as C5G7, containing mixed oxide (MOX) and UO2 fuel assemblies, to show the effectiveness of the proposed method. The results clearly indicate that the method is quite promising and acceptable.O. Safarzadeh, A. Minuchehr, A.S. ShiraniArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137239http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1372390306-45491873-2100http://dx.doi.org/10.1016/j.anucene.2013.06.011137240Resonance self-shielding calculation using subgroup method and ABC algorithmResonance self-shielding calculation using subgroup method and ABC algorithm1393-10-172015-01-07In this study, we propose a new method to optimize subgroup parameters for resonance self-shieldingcalculation. Our approach integrates the merits of both the subgroup method and ABC optimizationtechnique to effectively evaluate self-shielded resonance cross-sections. The ABC algorithm is used toobtain subgroup level in a way that guarantees reproduction of shielded effective cross sections in thesubgroup formulation. The temperature dependency of the cross-section is included in both subgrouplevel and subgroup weight. We used the conjugate gradient method based on the normal equations(CGNR) to evaluate the subgroup weights. An iteration technique is also used to consider the resonanceinterference. The proposed method is verified by analyzing Rowlands benchmark problems and Mosteller benchmark problems and comparing the obtained results with corresponding Monte Carlo solutions. The multiplication factor results show small errors and also good agreement.Omid Safarzadeh, A.S. Shirani, A. MinuchehrArticleProgress in Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137240http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1372400149-1970____-____http://dx.doi.org/10.1016/j.pnucene.2014.10.008137241Hybrid space–angle adaptivity for whole-core particle transport calculationsHybrid space–angle adaptivity for whole-core particle transport calculations1394-03-172015-06-07Adaptive refinement is a powerful method for efficiently solving physical problems. In this paper we present a new coupled space–angle adaptive algorithm for neutron transport calculations. The scheme is specifically employed for the solution of the integral form of transport equation based on the collision probability–response matrix method. The adaptive algorithm is started by first applying angular adaptivity and then projecting the solution to the spatial mesh refinement. A posteriori error estimate is derived by utilizing the flux gradient approach based on the net current leakage of nodes. A new approach is used to apply continuity of flux in the interface between nodes by escalating the order of spherical harmonics expansions of entrance response matrix to the same order of spherical harmonics expansions of outgoing angular flux at the neighboring node. Using an integral transport method within the node and refined space and angle variables, a new method for whole-core transport calculations is introduced. The validity of the developed adaptive strategy is assessed by a series of numerical experiments. Comparisons indicate that the space–angle adaptivity framework is capable of resulting acceptable solution with less number of the degrees of freedom (DOFs).Omid Safarzadeh, A.S. Shirani, A. MinuchehrArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137241http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1372410306-45491873-2100http://dx.doi.org/10.1016/j.anucene.2015.02.018137242An adaptive node refinement for particle transport techniques based on response matrix and collision probability methodsAn adaptive node refinement for particle transport techniques based on response matrix and collision probability methods1394-03-172015-06-07In this paper, we investigate and present an adaptive node refinement algorithm driven by an error estimator for particle transport techniques based on response matrix and collision probability methods. We represent the error estimator as a function of interface currents of the node. This indicator is used to drive a node refinement strategy for the spatial discretization. All nodes are derived from a common coarse node by hierarchic refinement. We assembled and coupled nodes from different sized nodes. Two problems featuring strong heterogeneity and highly transport streaming regime with strong flux gradients are used to test the adaptive refinement algorithm. The results show that the algorithm indeed bounds the flux error and achieves higher accuracy compared to uniform refinement.O. Safarzadeh, A.S. Shirani, A. MinuchehrArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137242http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1372420306-45491873-2100http://dx.doi.org/10.1016/j.anucene.2014.12.034137423Development of the triangle-based nodal algorithm for reconstructing pin power distributionsDevelopment of the triangle-based nodal algorithm for reconstructing pin power distributions1398-11-252020-02-14The fuel pin power is an essential parameter for increasing the safety and reliability features of the reactor. Although the high-order triangle-based polynomial expansion nodal (TPEN) algorithm had been suggested for incredible accuracy and computational speed in hexagonal core analyses, the pin power reconstruction of this algorithm is not developed and assessed. This paper presents the characteristics and performances of the TPEN algorithm for pin power reconstruction. The converged nodal information obtained from sweep between coarsemesh finite difference (CMFD) and TPEN methods is used to reconstruct the neutron flux distribution in a homogeneous fuel assembly. The modulation technique is used to obtain the heterogeneous distribution of power density. In this technique, the power density homogeneous distribution, calculated with the reconstructed neutron flux is multiplied by a form function. These functions are generated by DRAGON5. The results obtained by this algorithm are verified for various core configurations of a VVER-1000 reactor. The pin power factors show good agreement with the reference solution obtained by heterogeneous fine mesh finite element method. The largest and average relative errors found were of the order of 5 and 0.5 for TPEN, in a peripheral cell of a fuel element with the faces towards the region of the baffle/reflector. We also compared the results with those obtained from the nodal expansion method (NEM). The maximum and mean relative error of 10 and 1 are found by the NEM method.Omid SafarzadehArticleProgress in Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137423http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1374230149-1970____-____http://dx.doi.org/10.1016/j.pnucene.2020.103282137523Study the effects of various parameters on hydrogen production in the WWER1000/V446Study the effects of various parameters on hydrogen production in the WWER1000/V4461399-01-292020-04-17Hydrogen generation and its deflagration or explosion can damage the containment integrityduring severe accidents (SAs). In this paper, the total amount of the produced hydrogen duringStation Black-Out (SBO), SBO along with Large Break Loss of Coolant Accident (LBLOCA) andSBO along with Small Break Loss of Coolant accident (SBLOCA) for the WWER1000/V446reactor is estimated by the MELCOR code. Parametric analysis on the SA type, passive safetysystems, coating materials of the reactor pressure vessel (RPV), cavitys concrete type andcondition is discussed to determine their effects on the H2 production. The obtained resultsshowed that the SBO could produce more H2 (450 kg) compared with other SAs. Passive safetysystems increase the amount of the hydrogen production. In addition, CRBR concrete produceslow amount of the hydrogen compared with other concrete types.Gharari, Kazeminejad, Kojouri, Hedayat, Hassan-Vand, Nasiri, Omid SafarzadehArticleProgress in Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137523http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1375230149-1970____-____https://www.sciencedirect.com/science/article/pii/S0149197020301220137614Comparison of multi-point kinetics method with neutron diffusion equation and common point kinetics to investigate transient behavior in reactorsمقايسه سينتيک چند نقطهاي با معادله پخش نوترون و سينتيک نقطهاي متداول براي بررسي رفتار گذرا در راکتورها 1397-11-012019-01-21یکی از جنبههای مهم در طراحی و عملکرد یک راکتور هستهای، بررسی رفتار راکتور در طی حالات گذرا و شرایط غیر پایا است. برای این منظور، روشهای مختلفی برای تحلیل حالت گذرا ارائه شده است. حل مستقیم معادله پخش نوترون در حالت گذرا به همراه معادله غلظت مولدهای نوترون تاخیری یکی از روشهای دقیق ولی پرهزینه از نظر محاسباتی به حساب میاید. استفاده از این روشها در طراحی سیستم کنترل توان راکتورهای هستهای نیز موجب پیچیدگی کنترل کننده میشود که پیادهسازی ان را در عمل با مشکل مواجهه میکند. از اینرو، در طراحی سیستم کنترل راکتور هستهای معمولاً از سینتیک نقطهای استفاده میشود که تغییرات مکانی شار در نظر گرفته نمیشود. اخیراً سینتیک چند نقطهای برای کاستن این نقصان ارائه شده است. این روش در طراحی سیستم کنترل استفاده شده است اما دقت ان بررسی نشده است. این مقاله به مقایسه سینتیک چند نقطهای در تغییرات انی میله کنترل و مقایسه با سینتیک نقطهای متداول و حل مستقیم معادله پخش وابسته به زمان میپردازد. استخراج معادلات سینتیک چند نقطهای با استفاده از معادلات پخش دو گروهی انجام و از روش رانجی کوتا مرتبه 4 برای حل معادلات، استفاده شده است. نتایج نشانگر این است که، سینتیک چند نقطهای رفتار گذرا را با انحراف کمتری از سینتیک یک نقطهای محاسبه میکند. انحراف ناشی از تغییرات زیاد راکتیویته بیشتر از حالتی است که تغییرات راکتیویته ارام باشد. انحراف ناشی از تغییرات زیاد راکتیویته بیشتر از حالتی است که تغییرات راکتیویته ارام باشد. بنابراین سینتیک چند نقطه بیشتر برای بررسی تغییرات ارام و کند راکتیویته پیشنهاد میشود.Simin Mehrabi, Omid SafarzadehArticleفناوري و انرژي هسته ايژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137614http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137614____-________-____http://nucte.sbu.ac.ir/article/view/29068137628LBLOCA accident investigation using TRACE code in a VVER-1000 reactorبررسي حادثه LBLOCA با استفاده از کد TRACE در راکتور 1000-VVER1399-03-312020-06-20حادثه از دست رفتن خنککننده ناشی از کاهش حجم سیال خنککننده مدار اول است. عامل مستقیم این حادثه، خرابی یا خستگی مکانیکی ماده تشکیلدهنده اجزای مدار اول در هنگام عملکرد نیروگاه است. این حادثه که یک حادثهی مبنای طرح است و عامل مهمی در ارزیابی ایمنی نیروگاه هستهای است. درصورتیکه شکست در خط لوله اصلی مدار اول با قطر بیش از 25 درصد سطح مقطع خط لوله رخ دهد، به ان شکست بزرگ اطلاق میشود. در این مقاله، حادثه فوق با قطر شکست 850 میلیمتر، با استفاده از کد TRACE در یک راکتور 1000VVER- مدلسازی و تحلیل شده است. کد TRACE به صورت خاص برای حادثه از دسترفتن سیال خنک کننده طراحی شده است. با این تحلیل میتوان به جای فرضیات محافظه کارانه در ارزیابی ایمنی راکتور براورد دقیقی از ایمنی راکتور داشت و ملاحظات اقتصادی قابل توجهی بهدست اورد. در پایان، نتایج بهدست امده از کد TRACE با دادههای گزارش نهایی تحلیل ایمنی نیروگاه و همچنین نتایج تحقیقات پیشین مبتنی بر RELAP مقایسه شده است. نتایج نشانگر دقت کدTRACE در مدلسازی حادثه شکست بزرگ است.S. Ekbatani-Amlashi, A.S. Shirani, Omid SafarzadehArticleعلوم و فنون هسته ايژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137628http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1376281735-1871____-____https://jonsat.aeoi.org.ir/article_1094.html137771Core analysis of accident tolerant fuel cladding for SMART reactor under normal operation and rod ejection accident using DRAGON and PARCSCore analysis of accident tolerant fuel cladding for SMART reactor under normal operation and rod ejection accident using DRAGON and PARCS1399-06-252020-09-15There has been a deep interest in trying tofind better-performing fuel clad motivated by the desire todecrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. Inthis study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting conceptfor improving reactor safety is investigated for SMART modular reactor. These reactors have less pro-duction cost, short construction time, better safety and higher power density. The SiC and FeCrAl ma-terials are considered as the most potential candidate for ATF cladding, and the results are comparedwith Zircaloy cladding material from reactor physics point of view. In this paper, the calculations areperformed by generating PMAX library by DRAGON lattice physics code to be used for further reactorcore analysis by PARCS code. The differential and integral worth of control and safety rods, reactivitycoefficient, power and temperature distributions, and boric acid concentration during the cycle areanalyzed and compared from the conventional fuel cladding. The rod ejection accident (REA) is alsoperformed to study how the power changed in response to presence of the ATF cladding in the reactorcore. The key quantitativefinding can be summarized as: 20C (3) decrease in average fuel temper-ature, 33 pcm (3) increase in integral rod worth and cycle length, 1.26 pcm/C (50) and 1.05 pcm/C(16) increase in reactivity coefficient of fuel and moderator, respectively.Omid Safarzadeh, A. Pourrostam, S. TalebiArticle Nuclear Engineering and Technologyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137771http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1377711738-57331738-5733http://dx.doi.org/10.1016/j.net.2020.08.025137780Effects of the move towards Gen IV reactors in capacity expansion planning by total generation cost and environmental impact optimizationEffects of the move towards Gen IV reactors in capacity expansion planning by total generation cost and environmental impact 1399-06-242020-09-14Nowadays, it is necessary to accelerate the construction of new power plant in face of rising energy demand in such a way that the electricity will be generated at the lowest cost while reducing emissions caused by that generation. The expansion planning is one of the most important issues in electricity management. Nuclear energy comes forward with the low-carbon technology and increasing competitiveness to expand the share of generated energy by introducing Gen IV reactors. In this paper, the generation expansion planning of these new Gen reactors is investigated using the WASP software. Iran power grid is selected as a case of study. We present a comparison of the twenty-one year perspective on the future with the development of (1) traditional thermal power plants and Gen II reactors, (2) Gen III+ reactors with traditional thermal power plants, (3) Gen IV reactors and traditional thermal power plants, (4) Gen III+ reactors and the new generation of the thermal power plant, (5) the new generation of thermal power plants and the Gen IV reactors. The results show that the Gen IV reactors have the most developing among other types of power plants leading to reduce the operating costs and emissions. The obtained results show that the use of new Gen of combined cycle power plant and Gen IV reactors make the emissions and cost to be reduced to 16 and 72 of Gen II NPPs and traditional thermal power plants, respectively.Ali Bamshad, Omid SafarzadehArticle Nuclear Engineering and Technologyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=137780http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1377801738-57331738-5733http://dx.doi.org/10.1016/j.net.2020.09.005147983The station blackout accident in a dual-cooled annular fuel of a VVER-1000 reactor with application of portable pumps for mitigating the accidentThe station blackout accident in a dual-cooled annular fuel of a VVER-1000 reactor with application of portable pumps for mitigating the accident1399-08-212020-11-11The station blackout accident is clearly a very serious problem that jeopardizes the safety barrier ofnuclear reactor and ascends the core integrity steeply. The dual-cooled annular fuel has been introduced to boost the generated power density and safety margins. The fuel is cooled both internally and externally by allowing the liquid to flow on inner and outer heated surfaces. The fuel peak temperature can be reduced substantially by decreasing the fuel rod cladding heat flux leading to an increment on the DNBR. In this paper, a dual cooled annular fuel a VVER-1000 is investigated in this accident scenarios. The deployment of portable pumps for injecting coolant to dual cooled annular fuel core to mitigate this accident is also considered. The effect of accident tolerant fuel rod cladding is also investigated. The RELAP5 model is used to model the annular fuel to perform the analysis. The results are compared to traditional solid fuel geometry.A.M. Mehri, S. Talebi, Omid SafarzadehArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=147983http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1479830306-45491873-2100https://www.sciencedirect.com/science/article/pii/S0306454920306605148130Loading pattern optimization of a PWR reactor fuel assemblies using PESA-II algorithm by considering neutronic and thermal-hydraulic parametersبهينهسازي چيدمان مجتمعهاي سوخت يک راکتور PWR با استفاده از الگوريتم PESA-II و در نظر گرفتن پارامترهاي نوتروني و ترموهيدروليکي1399-10-012020-12-21روشهای بهینهسازی چندهدفه به صورت گستردهای در زمینههای مختلف علوم و مهندسی مورد توجه قرار گرفته است. در این مقاله، از الگوریتم بهینهسازی چندهدفه برای طراحی ارایش بهینه مجتمعهای سوخت قلب راکتور KWU با اهداف مسطح کردن توزیع توان در قلب راکتور و افزایش ضریب تکثیر استفاده شده است. همچنین، تحلیل ترموهیدرولیکی به منظور بررسی ارایش جدید طی یک حادثه فرضی کاهش جریان خنککننده انجام شده است. برای این منظور از ترکیب الگوریتم PESA-II با کد بسط یافته نودال (برای انجام محاسبات نوترونیک) و کد COBRA-EN (برای انجام محاسبات ترموهیدرولیک) استفاده شده است. نتایج نشان دادند که با جابهجایی چینش مجتمعهای سوخت پارامترهای نوترونیکی شامل ضریب تکثیر (افزایش از 0047/1 به 0058/1) و مسطح شدن توان بهبود یافتند. همچنین برای چینش جدید بررسی شد که در صورت وقوع حادثه فرضی کاهش جریان خنک کننده تا 15 درصد، راکتور همچنان میتواند در شرایط ایمن به کار خود ادامه دهد. علاوه بر این نتایج تاکید بر کارایی بهتر الگوریتم PESA-II نسبت به الگوریتم SPEA دارند.Omid Safarzadeh, B. Salmasian, R. GharariArticleعلوم و فنون هسته ايژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=148130http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1481301735-1871____-____https://jonsat.nstri.ir/article_1160.html148293Full-core reactor physics analysis for accident tolerant cladding in a VVER-1000 reactorFull-core reactor physics analysis for accident tolerant cladding in a VVER-1000 reactor1399-11-182021-02-06Advanced accident tolerant cladding materials have brought up the potential to delay the deleterious consequences of loos of coolant accidents related to slowing down hydrogen formation from reaction of zirconium with steam in order to minimize the additional heat generation and improve fuel and cladding retention of fission products. The performance improvement offered by these advanced materials may expand the operating envelope of existing light water reactors. This paper examines the neutronic performance of the VVER-1000 light water reactor for the application of accident tolerant cladding in order to realize the endurance of severe accident conditions. This study includes a detailed analysis of the control rod worth, reactivity coefficient, fuel cycle length, and power distribution for three accident tolerant cladding candidates of Ferritic-based alloy (FeCrAl), silicon carbide (SiC), and chromium coating application on zirconium claddings (ZrCr). The analysis was performed using diffusion-based core code PARCS, and lattice physics code DRAGON, including a developed package for regeneration of cross-sections in PMAX format.The reactor performance analysis indicates that the SiC cladding would have improved performance in terms of fuel cycle length, fuel and moderator temperature reactivity coefficients, and control rod worth. The cycle length would significantly decrease in magnitude for FeCeAl. Therefore, decreasing the cladding thickness by half and increasing the fuel enrichment by factor of 1.1875 made it possible to satisfy the required cycle length. A higher enrichment is also necessary for ZrCr cladding to increase the fuel burnup limits at nominal operating conditions.Omid Safarzadeh, M. QaraniArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=148293http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1482930306-45491873-2100https://www.sciencedirect.com/science/article/pii/S0306454921000396158428A fractional PID controller based on fractional point kinetic model and particle swarm optimization for power regulation of SMART reactorA fractional PID controller based on fractional point kinetic model and particle swarm optimization for power regulation of SMART reactor1399-12-222021-03-12The small modular reactor tends to drive down the price of electricity and heat, installed far from the national power grid. In order to provide the active power balance in small networks in fluctuating power demand, the output power of the reactor should be regulated. In this paper, a reactor core model for Korean integral-type small reactor, SMART, is proposed and verified in a rod ejection accident. A fractional controller intended to regulate the reactor power to chase the power demand. The particle swarm optimization has been carried out to minimize a certain cost function for step response of the original nonlinear plant. Simulation results show the excellent tracking of the desired output with practical control rod velocity and reactivity. The framework provided for the design of the FOPID shows the robust stability in the Nichols chart.Omid Safarzadeh, Omid Noori-kalkhoranArticleNuclear Engineering and Designژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=158428http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1584280029-5493____-____https://www.sciencedirect.com/science/article/abs/pii/S0029549321000893158667Full scope simulation of VVER-1000 blowdown source and containment pressurization in a LBLOCA by parallel coupling of TRACE and CONTAINFull scope simulation of VVER-1000 blowdown source and containment pressurization in a LBLOCA by parallel coupling of TRACE and CONTAIN1400-05-072021-07-29Nuclear power plants containment plays an important role as last-defined barrier in defense in depth approach against the release of radioactive material to the environment. In this study, a parallel processing couple has been developed to full scope analysis of blowdown source and containment pressurization parameters in a LBLOCA accident. To achieve this goal, primary and secondary loops of a VVER-1000/V446 were first simulated in TRACE V5.0 and steady-state results have been validated against reference data. The second step deals with containment simulation in CONTAIN 2.0 with new modified 30-cells models. A parallel processing interface was developed in MATLAB to couple TRACE and CONTAIN in the break point. Containment average pressure has been fed back to TRACE as forcing function of blowdown source in each time step during pressurization phase (coupling point). Finally, results of blowdown and containment pressurization have been validated against final safety analysis report (FSAR). Results of simulation confirm that the maximum containment pressure can reach 0.36 MPa and 0.395 MPa for this study and FSAR respectively that are lower than the maximum design absolute pressure of 0.46 MPa, so containment maintains its integrity during this accident. Temperature profiles of different control volumes inside containment during accident follow the FSAR profiles in terms of shape and value that show the ability of developed parallel coupling to full scope simulation of accidents accurately.Omid Safarzadeh, Omid Noori-kalkhoran, Massimiliano Gei, Lorenzo Morini, Rohollah AhangariArticleProgress in Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=158667http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1586670149-1970____-____https://www.sciencedirect.com/science/article/pii/S0149197021002602158901Application of metaheuristics optimization in fuel rod design: A case study for helium charging pressureApplication of metaheuristics optimization in fuel rod design: A case study for helium charging pressure1400-07-192021-10-11In this paper, it is suggested that the design parameters of the fuel rod can be optimized by using a metaheuristics algorithm so that the fuel rod is entirely safe in its normal operation and has a more efficient performance than usual. We applied Artificial Bee Colony (ABC) in fuel rod design. The fuel burnup, irradiation history, and operation data of the heavily burned fuel rod are obtained for typical VVER-1000 NPP by modeling core cycle design data from the first to the fourth cycle in the KASKAD package. This package intended for core calculation of VVER reactors consists of several primary modules such as BIPR-7A for assembly-wise core calculation, PERMAK-A for pin-by-pin core calculation, and TVS-M for macroscopic cross-section generation. The conventional initial gas pressure in the VVER reactor is 2 MPa. In this paper, considering the effects of gap gas pressure on the heat distribution of the fuel rod, the objective function has been selected to reduce the maximum temperature and the average temperature of the fuel pellet while maintaining safety standards. The optimal pressure obtained in this paper for this particular reactor with a specific power history is 0.467 MPa which is entirely safe in terms of safety throughout the reactor operation. It reduces the average fuel temperature during the process. Furthermore, the core calculations show that using this optimal pressure increases the quantity of energy produced. Fuel burnup increases at optimum pressure with an overall decrease in fuel temperature.Omid Safarzadeh, M. Ghasabian, S. TalebiArticleProgress in Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=158901http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1589010149-1970____-____https://www.sciencedirect.com/science/article/abs/pii/S0149197021003425159261Evaluation of advanced accumulator in a VVER-1000 reactor in loss of coolant accidentEvaluation of advanced accumulator in a VVER-1000 reactor in loss of coolant accident1401-04-182022-07-09The advanced accumulator is a passive system which gradually decreases the high flow into a low flow. The role of the fluidic device is to obtain an additional safety margin by increasing golden time to cope in the LOCA. In this paper, the effect of using the safety injection tank and the fluidic device, called advanced accumulator, in a wide spectrum of LOCA break sizes without emergency core cooling systems (ECCS) of a VVER-1000 is evaluated. Firstly, the main component of primary and secondary loops of Bushehr nuclear power plant (BNPP) is modeled by RELAP5. Secondly, the performance of the model is assessed in steady-state and LOCA conditions by comparing the obtained results with the plant’s FSAR. Thirdly, a model is developed in RELAP5 to simulate the behavior of advanced accumulator and validated by blowdown test in a test facility and MARS-KS code. Finally, the performance of advanced accumulator in the spectrum of 200–300 mm break sizes of LOCA without ECCS. In this range of breaks, the core damage times (CDT) developed from peak cladding temperature (PCT) of 1204 C have been calculated. In accordance with the results, the advanced accumulators can increase the CDT in the whole range of breaks.M. Pouresgandar, Omid Safarzadeh, S. TalebiArticleAnnals of Nuclear Energyژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=159261http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1592610306-45491873-2100https://www.sciencedirect.com/science/article/pii/S0306454922000238159386Estimation of effective B-fraction of the VVER-1000 reactor in terms of exposure using DRAGON5/DONJON5Estimation of effective B-fraction of the VVER-1000 reactor in terms of exposure using DRAGON5/DONJON51400-12-062022-02-25The effective β-fraction has a key role in the dynamic response of the reactor. This study aims to assess the suitability and accuracy of the detailed models of DRAGON5 and DONJON5 code for estimation of the effective fraction of delayed neutron for the VVER-1000 reactor core. DRAGON5 is adopted to homogenize and condense lattice physics constants of fuel assemblies during fuel burnup, followed by DONJON5, which is used to calculate forward and adjoint flux profiles on the reactor core geometry. A thermal-hydraulic subroutine is developed for VVER-1000 reactor hollow fuel pellets to embody the reactivity feedback raised by changing the reactor power profile. The effective β-fraction is evaluated for each fissile and fertile isotopes in terms of fuel burnup. The results of the coupling scheme are evaluated using the KASKAD code package of Bushehr NPP-I (BNPP-I). The results indicate that the use of SHI and SYBILT modules of DRAGON5 are essential to achieve reasonably precise resolution.Omid Safarzadeh, Farahnaz SaadatianArticleRadiation Physics and Engineeringژورنالي Full paper 1http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=159386http://research.shahed.ac.ir/WSR/WebPages/Report/PaperView.aspx?PaperID=1593862645-63972645-5188http://rpe.kntu.ac.ir/article_145172.html