Publications

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Books .

1. Atomic physics manual experiments for 2nd stage physics students

2. Electronics experiments for 3rd stage physics students.

Journal Papers:

1. Jamal M. Rashid Abda," Steady State Temperature Distribution Calculation in BN-350 Reactor Fuel Rod", J of Zankoy Sulaimani part A,V5 , no.1,43-49(2002).

Abstract : The steady state temperature distribution in one dimension for BN-350 reactor fuel rod has been calculated via solving heat conduction equation numerically . The calculations covers the case of constant and variable heat transfer coefficient considering different heat flux and inlet reactor core temperature.

The obtained results at average heat flux and core inlet temperature have been compared with reported results, it is found that these results could be a good references for reactor designing and for reactor safety processes when compared with higher values of material melting points.

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2. Jamal M. Rashid Abda,"TRansient temperature distribution calculation for BN-350 fuel rod", J of Dohuk university,1999.

Abstract : The temperature distribution as a function of time in a fuel rod of BN-350 fast reactor has been studied by establishing a computer program program that depends on the solving of the heat conduction equation numerically.Two conditions have been studied.The first is the distribution of temperature with time for Qmax =198.84(cal/cm³.sec) and with the flow of the sodium coolant, and the second case is for no flow of the coolant and with Qmax reduced to 186.68 cal/cm³.sec for two cases; when the control rods where up and when down.

Nuclear safety condition were considered in these calculations when dealing with temperatures greater than melting points of fuel and clad.The program depends on the finite difference technique to solve the conduction equation at each point at which the thermal conductivity and heat transfer coefficient where considered variables.

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3.Hamid Al-Bahily, Jamal M. Rashid Abda and marewan Ahmed Rashid,"The remainder coolant Quantity limitation during unsteadily flow in high pressure vessel of PWR-440 Mw", Iraqi J. Sci.,Vol. 45c, No.1, .(2004).

Abstract : Dynamic studying has been made in this research for the behavior of residual quantity at different times in high - pressure vessel of 400Mw reactor at sudden accident leakage in first cooling loop. Pore diameter leakage effect has been studied with its location, and coolant temperature inner the first cooling loop before leakage process.

Experiment equation was used for remainder coolant quantity calculation inner high-pressure vessel as a function of time depending on relative diameter and height of relative flux and coolant temperature.The research closed with an important conclusions which could contributed in construction of nuclear safety systems for reactors of type PWR with power of 440Mw.

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4. I have also three researches in IAEC in which they were forbidden to be published, 1988-1990 منظمة الطاقة الذرية العراقية.

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5. Jamal M.R.Abda and Mazin M. Elias"Calculation of dynamic coefficients for high energetic positrons implanted into a semi-crystalline polymer target film, Diala Journal for Pure Sciences.Vol: 12 No:3 , ISSN: 2222-8373, July 2016

Abstract: In this work absorbed , backscattering and transmitted probability coefficients has been calculated via Monte Carlo simulation technique PsMCS ( positronium Monte Carlo Computer Simulation ) for a high energetic positrons implanted normally into a semi finite Polytetrafluoroethylene target ( PTFE ) with its two components. Comparison with available references yields good quantitative agreement for dynamics factors.

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6. Jamal M. Rashid Abda,"Monte Carlo dynamic transmission calculations for high energetic positrons implanted into a semi-crystalline polymer target film Polytetrafluoroethylene),IJSR - INTERNATIONAL JOURNAL OF SCIENTIFIC RESEARCH,Volume : 5 | Issue : 1 | • ISSN No 2277 - 8179, Jan 2016

Abstract:

In this work, backscattering and transmitted probability coefficients with mean penetrating depth has been calculated via Monte Carlo simulation technique PsMCS ( positronium Monte – Carlo Computer Simulation ) for positrons with different energies starting by high energies and finishing with very slow energy ( from 0.15MeV to 70 eV) ,implanted normally and with different angles, into a semi finite Polytetrafluoroethylene target ( PTFE ). Comparison with available references yields good quantitative agreement for dynamics factors.

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7. Jamal M. Rashid Abda and Abdul Karim Thamir Mohammed,"Neutronic and Thermal – Hydraulic Calculations for The Conversion of BN-350, 1^st International Conference of Natural Science CHARMO University Chamchamal, ISSN 2520-5749,ICNS-2016, page 109, 11-12th july 2016.

Abstract:

This paper represents the neutronic and thermal-hydraulic calculations for the conversion of BN-350 from 27% to 7% enriched Uranium fuel elements. Mixed enrichment cores has been studied where low enrichment uranium (LEU) cores fresh fuel elements substitute gradually the high enrichment uranium (HEU) depleted fuel elements in the equilibrium core. Thermal hydraulic calculations have been carried out to determine changes in the characteristics of the converted reactor during steady state conditions and transient response to a coolant flow loss (LOCA).

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8. Jamal M. Rashid Abda, "Transmission and Backscatter Coefficients of 1 to 160 keV Positrons incident on a semi-crystalline polymer target film"; International Journal of Scientific & Engineering Research, Volume 8, Issue 3, ,ISSN 2229-5518, March-(2017).

Abstract:

The aim of this research is to investigate the behavior of the backscattering, transmitted and absorbed coefficients as a function of the positrons beam primary energy. The absorbed, backscattering and transmitted probability coefficients has been calculated and studied raphically via Monte Carlo simulation technique PsMCS ( Ppositronium Monte Carlo Computer Simulation ) for a high energetic positrons from 1 to 160 keV, implanted normally into a semi -crystalline Polytetrafluoroethylene target ( PTFE ) with it’s the two components of the target. Comparison with available references yields good quantitative agreement for dynamics factors.

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9. Jamal M. Rashid Abda, " Study of neutronic and temperature distribution during enrichment reduction for BN-350 reactor", Accepted to be published in Journal of University of Babylon.(1918).

Abstract:

This paper represents the neutronic and thermal-hydraulic calculations for the conversion of BN-350 from 27% to 7% enriched Uranium fuel elements. Mixed enrichment cores has been studied where low enrichment uranium (LEU) cores fresh fuel elements substitute gradually the high enrichment uranium (HEU) depleted fuel elements in the equilibrium core. Thermal hydraulic calculations have been carried out to determine changes in the characteristics of the converted reactor during steady state conditions and transient response to a coolant flow loss.

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