Pub Date : 2018-10-17DOI: 10.1504/IJNEST.2018.10016738
S. Pinem, T. M. Sembiring, T. Surbakti
The TRIGA 2000 Bandung (TRIGA 2000) is a TRIGA Mark II type reactor with the nominal thermal power of 2 MW. It is used as a back-up reactor for irradiation radioisotope targets to the local market. This present study describes the fuel type conversion, from TRIGA fuel rod type to the MTR fuel plate type, as a solution to the fuel shortage. The optimum silicide (U3Si2-Al) equilibrium core are obtained by using WIMSD/5B (cell calculations) and BATAN-FUEL and BATAN-3DIFF (core calculations) codes for static and kinetic core parameters. The core calculation results showed that the optimum equilibrium core configuration is 5×5 core grid position and the maximum thermal neutron fluxes in the centre and periphery core are 6.07×1013 neutron/cm2s and 1.41×1013 neutron/cm2s, respectively. It is concluded that TRIGA 2000 reactor can be designed with very satisfactory static parameters and inherent safety.
TRIGA 2000万隆(TRIGA 2000)是TRIGA Mark II型反应堆，标称热功率为2兆瓦。它被用作向当地市场照射放射性同位素目标的备用反应堆。本文描述了燃料类型的转换，从TRIGA燃料棒型到MTR燃料板型，作为解决燃料短缺的方法。采用WIMSD/5B(电池计算)和BATAN-FUEL和BATAN-3DIFF(核心计算)代码计算了硅化物(U3Si2-Al)的静态和动态核心参数，得到了最优的平衡核心。计算结果表明，堆芯的最佳平衡构型为5×5堆芯网格位置，堆芯中心和外围的最大热中子通量分别为6.07×1013 neutron/cm2s和1.41×1013 neutron/cm2s。结果表明，TRIGA 2000型反应堆的静态参数和固有安全性是非常令人满意的。
{"title":"Core conversion design study of TRIGA Mark 2000 Bandung using MTR plate type fuel element","authors":"S. Pinem, T. M. Sembiring, T. Surbakti","doi":"10.1504/IJNEST.2018.10016738","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10016738","url":null,"abstract":"The TRIGA 2000 Bandung (TRIGA 2000) is a TRIGA Mark II type reactor with the nominal thermal power of 2 MW. It is used as a back-up reactor for irradiation radioisotope targets to the local market. This present study describes the fuel type conversion, from TRIGA fuel rod type to the MTR fuel plate type, as a solution to the fuel shortage. The optimum silicide (U3Si2-Al) equilibrium core are obtained by using WIMSD/5B (cell calculations) and BATAN-FUEL and BATAN-3DIFF (core calculations) codes for static and kinetic core parameters. The core calculation results showed that the optimum equilibrium core configuration is 5×5 core grid position and the maximum thermal neutron fluxes in the centre and periphery core are 6.07×1013 neutron/cm2s and 1.41×1013 neutron/cm2s, respectively. It is concluded that TRIGA 2000 reactor can be designed with very satisfactory static parameters and inherent safety.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"222"},"PeriodicalIF":0.0,"publicationDate":"2018-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44156865","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-10-17DOI: 10.1504/IJNEST.2018.10016743
A. Mathur, S. Khan, V. Jagannathan, L. Thilagam, Suneet Singh
The LWR benchmark suite is a set of problems that was proposed for validating lattice calculations at pin and assembly level for long life fuel. The LEU and MOX fuel assembly designs consider fissile loading and gadolinium loading which are significantly higher than current fuel designs. In this paper, we describe the solution to this benchmark problem by the lattice burnup code VISWAM. This code solves the transport equation by the interface current method based on 2-D collision probability (2DCP). The incoming and outgoing angular fluxes at the pin-cell interfaces are expanded in terms of PN half space expansions, where N is the order of the expansion. We have considered three expansions, double P0, double P1 and double P2 for evaluating the escape and transmission probabilities. The higher order methods are more accurate where flux anisotropy is high.
{"title":"Analysis of long life LWR fuel benchmark by CP based interface current methods","authors":"A. Mathur, S. Khan, V. Jagannathan, L. Thilagam, Suneet Singh","doi":"10.1504/IJNEST.2018.10016743","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10016743","url":null,"abstract":"The LWR benchmark suite is a set of problems that was proposed for validating lattice calculations at pin and assembly level for long life fuel. The LEU and MOX fuel assembly designs consider fissile loading and gadolinium loading which are significantly higher than current fuel designs. In this paper, we describe the solution to this benchmark problem by the lattice burnup code VISWAM. This code solves the transport equation by the interface current method based on 2-D collision probability (2DCP). The incoming and outgoing angular fluxes at the pin-cell interfaces are expanded in terms of PN half space expansions, where N is the order of the expansion. We have considered three expansions, double P0, double P1 and double P2 for evaluating the escape and transmission probabilities. The higher order methods are more accurate where flux anisotropy is high.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"252"},"PeriodicalIF":0.0,"publicationDate":"2018-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48315168","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-10-17DOI: 10.1504/IJNEST.2018.10016742
Rogério Rivail Rodrigues, A. Z. Mesquita, D. A. Palma
The Nuclear Technology Development Centre (CDTN) is a research institute administered by the Brazilian Nuclear Energy Commission (CNEN) located in Belo Horizonte/Brazil. The CDTN is equipped with a Triga Mark I nuclear reactor with almost 60 years of operation. Most of its fuel elements are in the core since the first criticality, and corrosion may occur that threatens the integrity of spent fuel claddings. The reactor coolant must be treated and controlled in order to maintain its low electrical conductivity and pH close to neutrality, in order to minimise the corrosion of the reactor components, mainly the fuel elements. The objective of this work is to present the leak monitoring system developed for the verification of possible leaks in the Triga fuel elements based on the sipping test. A correlation was also developed to find the diameter of a hypothetical small cylindrical hole in the fuel cladding as a function of Cs-137 activity values to be measured.
核技术发展中心是一个由巴西核能委员会管理的研究机构，位于巴西贝洛奥里藏特。CDTN配备了一个运行了近60年的Triga Mark I核反应堆。自第一次临界以来，其大部分燃料元件都在堆芯中，可能发生腐蚀，威胁乏燃料包壳的完整性。必须对反应堆冷却剂进行处理和控制，以保持其低电导率和接近中性的pH值，从而最大限度地减少反应堆部件（主要是燃料元件）的腐蚀。本工作的目的是介绍为验证基于啜饮试验的Triga燃料元件中可能存在的泄漏而开发的泄漏监测系统。还开发了一种相关性，以找到燃料包壳中假设的小圆柱形孔的直径作为待测量的Cs-137活度值的函数。
{"title":"Designing a system to detect leaking in fuel elements in Brazilian Triga research reactor","authors":"Rogério Rivail Rodrigues, A. Z. Mesquita, D. A. Palma","doi":"10.1504/IJNEST.2018.10016742","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10016742","url":null,"abstract":"The Nuclear Technology Development Centre (CDTN) is a research institute administered by the Brazilian Nuclear Energy Commission (CNEN) located in Belo Horizonte/Brazil. The CDTN is equipped with a Triga Mark I nuclear reactor with almost 60 years of operation. Most of its fuel elements are in the core since the first criticality, and corrosion may occur that threatens the integrity of spent fuel claddings. The reactor coolant must be treated and controlled in order to maintain its low electrical conductivity and pH close to neutrality, in order to minimise the corrosion of the reactor components, mainly the fuel elements. The objective of this work is to present the leak monitoring system developed for the verification of possible leaks in the Triga fuel elements based on the sipping test. A correlation was also developed to find the diameter of a hypothetical small cylindrical hole in the fuel cladding as a function of Cs-137 activity values to be measured.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"239"},"PeriodicalIF":0.0,"publicationDate":"2018-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46077807","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-10-17DOI: 10.1504/IJNEST.2018.10016725
B. Salmasian, G. Ansarifar, S. M. Mirvakili
In this paper, modelling of the Tehran Research Reactor is done using Recurrent Neural Network (RNN) in Loss of Flow Accident (LOFA). TRANS code is calculated as training data mode for each of the scenarios. Supervised recurrent neural network is chosen for modelling and identification system, classified system data and appropriate parameters for modelling function of system have been chosen, then data is classified. In the next step, we choose variant networks to train and compare with each other. Next, an optimised network is chosen according to mean square error parameter and correlation among educational data from TRANS code and network output data. Finally, entrance data related to the unforeseen accident was entered to the system and the predicted results by model and output data of TRANS code were compared. Results demonstrate the appropriate conformity between extraction data of TRANS code and extraction data of the model, which shows appropriate function of the model.
{"title":"System identification of a research nuclear reactor versus loss of flow accident using recurrent neural network","authors":"B. Salmasian, G. Ansarifar, S. M. Mirvakili","doi":"10.1504/IJNEST.2018.10016725","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10016725","url":null,"abstract":"In this paper, modelling of the Tehran Research Reactor is done using Recurrent Neural Network (RNN) in Loss of Flow Accident (LOFA). TRANS code is calculated as training data mode for each of the scenarios. Supervised recurrent neural network is chosen for modelling and identification system, classified system data and appropriate parameters for modelling function of system have been chosen, then data is classified. In the next step, we choose variant networks to train and compare with each other. Next, an optimised network is chosen according to mean square error parameter and correlation among educational data from TRANS code and network output data. Finally, entrance data related to the unforeseen accident was entered to the system and the predicted results by model and output data of TRANS code were compared. Results demonstrate the appropriate conformity between extraction data of TRANS code and extraction data of the model, which shows appropriate function of the model.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"283"},"PeriodicalIF":0.0,"publicationDate":"2018-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44145023","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-10-17DOI: 10.1504/IJNEST.2018.10016734
K. S. Jassim, I. H. Kadhim, Nawras T. Shihab
In this study, the concentration of radon in water has been measured for 17 samples from different regions of Al-Kifl in the province of Babylon, Iraq. This study was performed using the electronic radon detector RADH2O, where the highest value was 0.199 Bq.L−1 and the lowest value was 0.0 Bq.L−1 and the effective dose for human exposure to radon average is 5.09*10−7 Sv.y−1. The results of radon concentrations and annual effective dose in all samples show no significant radiological risk for the inhabitants in the region of study. We have chosen this subject for the current study because of the importance of water in human life and living, and the lack of previous studies in this study area.
{"title":"Estimated radon concentration in drinking water samples for different regions of Hilla City, Iraq","authors":"K. S. Jassim, I. H. Kadhim, Nawras T. Shihab","doi":"10.1504/IJNEST.2018.10016734","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10016734","url":null,"abstract":"In this study, the concentration of radon in water has been measured for 17 samples from different regions of Al-Kifl in the province of Babylon, Iraq. This study was performed using the electronic radon detector RADH2O, where the highest value was 0.199 Bq.L−1 and the lowest value was 0.0 Bq.L−1 and the effective dose for human exposure to radon average is 5.09*10−7 Sv.y−1. The results of radon concentrations and annual effective dose in all samples show no significant radiological risk for the inhabitants in the region of study. We have chosen this subject for the current study because of the importance of water in human life and living, and the lack of previous studies in this study area.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"213"},"PeriodicalIF":0.0,"publicationDate":"2018-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48915176","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-08-27DOI: 10.1504/IJNEST.2018.094284
E. M. Borges, G. Sabundjian, F. D’Auria, A. Petruzzi
Owing to the occurrence of nuclear accidents, worldwide nuclear regulatory organisations included the analysis of accidents considered as design basis accidents - Loss of Coolant Accident (large and small-break, LBLOCA or SBLOCA) - in the safety analysis reports of nuclear facilities. In Brazil, the tool selected by the licensing authority, Comissao Nacional de Energia Nuclear (CNEN), is RELAP5 Code. The aim of this paper is the evaluation of the performance of the Emergency Core Cooling System (ECCS) of Angra 2 nuclear reactor during SBLOCA. In this study, the RELAP5 code and the Code Internal Assessment of Uncertainty (CIAU) were used to simulate and analyse the uncertainties of the results. The postulated accident is the SBLOCA in the hot leg connected to the ECCS described in the Final Safety Analysis Report of Angra 2 (FSAR/A2). The results from this study were satisfactory when compared with the FSAR/A2.
{"title":"Uncertainty calculation in small break LOCA in the emergency core cooling system connected to the hot leg of Angra 2 nuclear power plant","authors":"E. M. Borges, G. Sabundjian, F. D’Auria, A. Petruzzi","doi":"10.1504/IJNEST.2018.094284","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.094284","url":null,"abstract":"Owing to the occurrence of nuclear accidents, worldwide nuclear regulatory organisations included the analysis of accidents considered as design basis accidents - Loss of Coolant Accident (large and small-break, LBLOCA or SBLOCA) - in the safety analysis reports of nuclear facilities. In Brazil, the tool selected by the licensing authority, Comissao Nacional de Energia Nuclear (CNEN), is RELAP5 Code. The aim of this paper is the evaluation of the performance of the Emergency Core Cooling System (ECCS) of Angra 2 nuclear reactor during SBLOCA. In this study, the RELAP5 code and the Code Internal Assessment of Uncertainty (CIAU) were used to simulate and analyse the uncertainties of the results. The postulated accident is the SBLOCA in the hot leg connected to the ECCS described in the Final Safety Analysis Report of Angra 2 (FSAR/A2). The results from this study were satisfactory when compared with the FSAR/A2.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"139-160"},"PeriodicalIF":0.0,"publicationDate":"2018-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1504/IJNEST.2018.094284","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48090514","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-08-27DOI: 10.1504/ijnest.2018.10015405
J. Bahmani, B. Eslami, F. Jafari
For the ignition of p6Li fuel, two important challenges emerge: the losses of energy and the need for high-temperature electrons and ions. To provide favourable conditions, these problems should be reduced as much as possible. An approach for reducing ignition energy is proposed which is the use of proton beam in fast ignition with inertial confinement fusion (ICF). The protons can fuse with ions and produce additional energy. The determination of this energy enhancement and the investigation of influential parameters can have a significant role in increasing energy gain. The stopping power is examined. The produced excess energy and the total energy are estimated. The investigations indicate that total stopping power is a function of proton energy and it decreases with rising temperature. The additional energy depends on the beam energy and the electron temperature. The extra energy gain is considerable at low proton energy and high electron temperature.
{"title":"The enhancement of energy gain in a p6Li inertial fusion reactor by laser-driven protons","authors":"J. Bahmani, B. Eslami, F. Jafari","doi":"10.1504/ijnest.2018.10015405","DOIUrl":"https://doi.org/10.1504/ijnest.2018.10015405","url":null,"abstract":"For the ignition of p6Li fuel, two important challenges emerge: the losses of energy and the need for high-temperature electrons and ions. To provide favourable conditions, these problems should be reduced as much as possible. An approach for reducing ignition energy is proposed which is the use of proton beam in fast ignition with inertial confinement fusion (ICF). The protons can fuse with ions and produce additional energy. The determination of this energy enhancement and the investigation of influential parameters can have a significant role in increasing energy gain. The stopping power is examined. The produced excess energy and the total energy are estimated. The investigations indicate that total stopping power is a function of proton energy and it decreases with rising temperature. The additional energy depends on the beam energy and the electron temperature. The extra energy gain is considerable at low proton energy and high electron temperature.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"127"},"PeriodicalIF":0.0,"publicationDate":"2018-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48085008","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-08-27DOI: 10.1504/IJNEST.2018.10015404
K. Mehboob, M. S. Aljohani
The CHASNUPP - unit 1 is a 996 MWth intermediate type pressurised water reactor that began commercial operation in May 2000 in Pakistan. The CHNUPP-1 is a conventional two loop PWR operated by the Pakistan Atomic Energy Commission (PAEC). The expected amount of radiation exposure from the CHNUPP-1 R CHNUPP - unit 1 is simulated for hypothetical severe accidents. For this purpose, modelling and simulation has been carried out in MATLAB. A kinetic model has been developed and implemented in MATLAB to carry out the simulation of the release of radionuclides. The core and coolant activity of CHNUPP-1 is compared with the similar type reactor KORI-1. The developed model uses the Origen 2.2 core inventory as a subroutine. The coolant inventory has been evaluated with 0.25% fuel damage and compared with the KORI-1 reactor.
CHASNUPP - 1号机组是一座996兆瓦的中压水反应堆，于2000年5月在巴基斯坦开始商业运行。CHNUPP-1是由巴基斯坦原子能委员会(PAEC)操作的常规双环压水堆。在假设的严重事故中，模拟了CHNUPP-1 R CHNUPP-1机组的预期辐射暴露量。为此，在MATLAB中进行了建模和仿真。在MATLAB中开发并实现了一个动力学模型，对放射性核素的释放过程进行了模拟。将CHNUPP-1堆芯和冷却剂活性与KORI-1进行了比较。所开发的模型使用Origen 2.2核心库存作为子例程。该冷却剂库存以0.25%的燃料损坏进行了评估，并与KORI-1反应堆进行了比较。
{"title":"Estimation of radioactivity released from CHASNUPP-1 nuclear power plant during loss of coolant accident","authors":"K. Mehboob, M. S. Aljohani","doi":"10.1504/IJNEST.2018.10015404","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10015404","url":null,"abstract":"The CHASNUPP - unit 1 is a 996 MWth intermediate type pressurised water reactor that began commercial operation in May 2000 in Pakistan. The CHNUPP-1 is a conventional two loop PWR operated by the Pakistan Atomic Energy Commission (PAEC). The expected amount of radiation exposure from the CHNUPP-1 R CHNUPP - unit 1 is simulated for hypothetical severe accidents. For this purpose, modelling and simulation has been carried out in MATLAB. A kinetic model has been developed and implemented in MATLAB to carry out the simulation of the release of radionuclides. The core and coolant activity of CHNUPP-1 is compared with the similar type reactor KORI-1. The developed model uses the Origen 2.2 core inventory as a subroutine. The coolant inventory has been evaluated with 0.25% fuel damage and compared with the KORI-1 reactor.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"111"},"PeriodicalIF":0.0,"publicationDate":"2018-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41975187","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-08-27DOI: 10.1504/IJNEST.2018.10015408
B. B. Filho, E. Cabral, A. Barroso
The International Association for the Properties of Water and Steam (IAPWS) develops formulations for the calculation of thermophysical properties of water as a function of different combinations of temperature, density, pressure, enthalpy, and entropy. These properties are useful for scientists and nuclear, chemical, and mechanical engineers who analyse experimental data or are involved with projects and equipment development, like heat exchangers, turbines, or nuclear power reactors. The IAPWS-95 formulation solves the fundamental equation of Helmholtz free energy as a function of temperature and density. This paper gives a description of how these equations are solved and exemplifies the use of a package developed for the free platform R. The IAPWS95 package was developed to help users to get access to the IAPWS-95 formulation in a free software environment which is growing exponentially. Transport properties were programmed using other IAPWS releases. The examples consider the uncertainty analysis of thermal parameters of a nuclear power reactor and the preparation of tables and graphs of water properties.
{"title":"An R-package for water and steam properties for scientific and general use","authors":"B. B. Filho, E. Cabral, A. Barroso","doi":"10.1504/IJNEST.2018.10015408","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10015408","url":null,"abstract":"The International Association for the Properties of Water and Steam (IAPWS) develops formulations for the calculation of thermophysical properties of water as a function of different combinations of temperature, density, pressure, enthalpy, and entropy. These properties are useful for scientists and nuclear, chemical, and mechanical engineers who analyse experimental data or are involved with projects and equipment development, like heat exchangers, turbines, or nuclear power reactors. The IAPWS-95 formulation solves the fundamental equation of Helmholtz free energy as a function of temperature and density. This paper gives a description of how these equations are solved and exemplifies the use of a package developed for the free platform R. The IAPWS95 package was developed to help users to get access to the IAPWS-95 formulation in a free software environment which is growing exponentially. Transport properties were programmed using other IAPWS releases. The examples consider the uncertainty analysis of thermal parameters of a nuclear power reactor and the preparation of tables and graphs of water properties.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"172"},"PeriodicalIF":0.0,"publicationDate":"2018-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41703677","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-08-27DOI: 10.1504/IJNEST.2018.10015407
Ned Xoubi, A. Y. Soliman
In this work, a full core, three-dimensional, multi-group model of VVER-1000 reactor core is developed using specifications of the Schulz benchmark. This paper presents a new 3-D full core solution, and simulates the neutronic behaviour of the VVER-1000 reactor by predicting the system criticality, power distribution, and the neutron flux distribution. Multi-group constants are applied to the COMSOL model to perform neutronics calculations using finite element method with adaptive mesh refinement. The study found that the calculated effective multiplication factor (keff) compares well with the reference value. Furthermore, the fission rates 3D power distributions and axially averaged 2D power distribution are in good agreements with reported reference results. The thermal neutron spectrum is also calculated by the COMSOL model as presented in this paper. This study allows us to validate the COMSOL calculation schemes for VVER-type reactors and to compare our solutions with reference solutions at steady state.
{"title":"Validating COMSOL multiphysics for VVER-1000 whole-core-steady-state via AER benchmark problem","authors":"Ned Xoubi, A. Y. Soliman","doi":"10.1504/IJNEST.2018.10015407","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10015407","url":null,"abstract":"In this work, a full core, three-dimensional, multi-group model of VVER-1000 reactor core is developed using specifications of the Schulz benchmark. This paper presents a new 3-D full core solution, and simulates the neutronic behaviour of the VVER-1000 reactor by predicting the system criticality, power distribution, and the neutron flux distribution. Multi-group constants are applied to the COMSOL model to perform neutronics calculations using finite element method with adaptive mesh refinement. The study found that the calculated effective multiplication factor (keff) compares well with the reference value. Furthermore, the fission rates 3D power distributions and axially averaged 2D power distribution are in good agreements with reported reference results. The thermal neutron spectrum is also calculated by the COMSOL model as presented in this paper. This study allows us to validate the COMSOL calculation schemes for VVER-type reactors and to compare our solutions with reference solutions at steady state.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"161"},"PeriodicalIF":0.0,"publicationDate":"2018-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49081284","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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