Amr Abdelhady | Nuclear Research Center, Atomic Energy Authority (original) (raw)

Papers by Amr Abdelhady

SSRN, 2023

One of the main studies that must be conducted before installing or building a radioisotopes prod... more One of the main studies that must be conducted before installing or building a radioisotopes production factory (RPF) is the radiological effect of the accidental releasing of the radioisotopes on the personnel and the environment surrounding the factory. The expected dose levels and the radioactive contamination areas must be determined to prepare the proper countermeasure after the hypothetical accident. In this study, the radioisotopes production factory is built inside a nuclear research center for production of 131 I and 99 Mo from irradiated low enrichment uranium (LEU) plates. It contains hot cells for extracting and preparing the 131 I and 99 Mo after irradiating the LEU plates inside the reactor core. The study focuses on evaluating the dose levels around the RPF after the accidental release of 131 I due to malfunction during the 131 I preparation process inside the iodine hot cell. Conservatively, the maximum amount of the 131 I release would be considered to calculate the maximum dose received by the personnel after the accident. The radiation dose levels around the RPF would be computed using GENII-2 code taking the site meteorological data into consideration. The results show that maximum dose would be located at distances from 800 to 5500 m from the RPF building at the direction ranging between 140 o and 200 o. The result shows that the maximum dose would be more than the permissible limit for the worker. The study ensures the importance of instant distributing the potassium iodide tablets to the persons who locate around the RPF building after the accident.

Nuclear science and engineering, 2023

The transfer of nuclear spent fuel from the reactor storage pool to dry storage or for reprocessi... more The transfer of nuclear spent fuel from the reactor storage pool to dry storage or for reprocessing or final disposition requires information about its isotopic composition, decay heat, and other thermomechanical properties. The spent nuclear fuel assembly of a typical advanced pressurized water reactor, AP-1000, was characterized using the Monte Carlo MCNPX code and SCALE/ORIGEN code. The simulation of operational history started from the operation of the first fresh core for an average fuel assembly with certain physical isotopic parameters until 25 GWd/tonne U discharge burnup. The analysis considered the calculations of the radionuclide inventories, activity, neutron emission spectrum, gamma-ray emission spectrum, and decay power after 700 effective full power days and for post different time ranges until a 1 million-year cooling period. The comparison of some results of the two codes showed small differences due to the consideration of the continuous-energy variation for neutrons in the MCNPX code and the discrete energy assumption in the SCALE/ ORIGEN code.

Loss of coolant accident (LOCA) is one of the anticipated accidents that may occur in the open po... more Loss of coolant accident (LOCA) is one of the anticipated accidents that may occur in the open pool type reactor. LOCA results in decreasing the water level above the reactor core and consequently may cause an increase in the radiation dose rate level. Determining the radiation dose rate levels around the reactor building due to the LOCA is the objective of this study. KENO-VI code was used to model the reactor building structure. ORIGEN-S code was used to evaluate the neutron and the photon spectrums and their intensities after the reactor shutdown assuming the maximum burnup of the discharge for conservative calculation. MAVRIC code was used to calculate the dose rate distributions around the reactor during LOCA. Then, the distances from the reactor which correspond to the permissible dose rate limit were determined along the water level drop. The study focused on determining the water level above the reactor core, 144 cm, that achieves a dose rate lower than or equal to the permissible limit for any position around the reactor building.

KERNTECHNIK, 2022

Production of high specific 60 Co activity in nuclear research reactor needs long time of irradia... more Production of high specific 60 Co activity in nuclear research reactor needs long time of irradiation to reach the proper consumer demands. Cobalt device is a fixed experiment inside the core of the second research reactor (ETRR-2) that designed for production of 60 Co with activity of 50,000 Ci. Cobalt device is assembly of 16 pins of aluminum cans that filling with the cobalt foils. Utilizing the programs that depend on the Monte Carlo method to estimate 60 Co activity is the objective of this study. MCNP6, KENO-VI, and MAVRIC codes would be used to evaluate the time that the cobalt device has been spent inside the reactor core for reaching the proper 60 Co activity considering the continuous irradiation as well as the discrete irradiation. The calculations show that nearly one year of continuous neutron irradiation is sufficient for reaching the required activity. In case of discrete irradiation, the time required for reaching the required activity would be more than a year depending on the reactor operation period.

ASME Journal of nuclear engineering and radiation science, 2021

The objective of this study is suggesting a new method for gamma irradiating experiment using spe... more The objective of this study is suggesting a new method for gamma irradiating experiment using spent fuel elements (SFE). An irradiation device was developed to contain the sample during the gamma irradiation process under water in spent fuel storage pool. The maximum burned fuel elements were used as a source of gamma rays with intensity and energy spectrum depend on the decay time from the reactor core withdrawal. A SFE configuration was introduced to verify an ideal condition of irradiating the sample with uniform gamma flux. The gamma dose rate inside the sample chamber was calculated using SCALE/MAVRIC code for the SFEs configuration source at decay times ranging between 0 to 5 year. A relationship was introduced to obtain the required accumulated dose depending on choosing the proper SFE configuration with the proper irradiation time.

Ssrn electronic journal , 2022

The objective of this work is design a shielded cask to transport the radioactive 99 Mo column fr... more The objective of this work is design a shielded cask to transport the radioactive 99 Mo column from the radioisotopes facility to the consumers. The cask must be designed to confine the column activity to decrease the dose rate outside the cask to the permissible limits. A different cask designs will be introduced to cover the 99 Mo activities variety depending on the consumer's demands. An available cask was evaluated from the radiological viewpoint to determine its capacity for confinement 99 Mo column of activity of 1 Ci. The developed shielded casks can contain activities ranging from 1 cCi to 100 Ci. MAVRIC module from SCALE system would be used to develop the shielding and to evaluate the dose rate around the cask loaded with the different 99 Mo activities.

BRAZILIAN JOURNAL OF RADIATION SCIENCES, 2022

This study aims to model the radiological consequences that result from loss of coolant accident ... more This study aims to model the radiological consequences that result from loss of coolant accident in open pool
research reactor of 22 MW power. The loss of coolant accident results from rupture in the cooling systems of the
reactor and consequently results in decreasing the water level in the reactor pool. Decreasing the water level,
that represents the shielding material above the reactor core, results in increasing the dose rate in the top region
in the reactor pool as well as the reactor building. LOCA normally results in shutting down the reactor due to
the radioprotection system provided in the upper region of the reactor pool. A failure in shutdown systems of the
reactor during LOCA is rare but expected and it represents the worst case. So, evaluating LOCA from the
radiological point of view must be investigated for the safe shutdown mode as well as the failure of shutdown
case. The study also includes occurring LOCA for shutdown reactor taking the decay time into consideration.
ORIGEN-S module from scale system was used to estimate the delayed neutrons and gamma rays from the core
for the shutdown mode. KENO-VI module from Scale system was used to estimate the prompt neutrons and
gamma rays generated from the core for the failure case. Then, MAVRIC/Scale module was used to estimate the
dose rate at the top point of the reactor pool for the two cases during LOCA.
Studying the dose rate during LOCA aims to determine the water level that verifies the radiological safety limit
at the top point of the reactor pool for the two cases. The calculations introduced a relationship between the
water level that verifies the permissible radiological limit (WL) and the decay time (DT) in the range between 0 to
30 d.

JOURNAL OF NUCLEAR RESEARCH AND DEVELOPMENT, NO. 22, DECEMBER 2021, 2021

Storage of spent fuel elements needs a previous calculation to ensure verifying the criteria of s... more Storage of spent fuel elements needs a previous calculation to ensure verifying the criteria of storage from the nuclear criticality point of view. So, arrangements of spent fuel elements in the storage must result in a configuration characterized by an effective multiplication factor (k-eff) lower than 0.85. The spent fuel elements storage facility is an open pool connected with the main pool of the reactor via a transfer channel to facilitate transport of the irradiated fuel elements under water surface to assure a high radiological safety. The fuel element is a plate type for the material testing reactor (MTR), U 3 O 8 fuel with enrichment of 19.97 wt% in 235 U, in a matrix of Al. The storage pool was designed to store around 800 spent fuel elements as a temporary stage before transporting to a permanent storage. Basket and pool models, loaded with fresh fuel elements, were used in KENO-VI criticality code included in SCALE6.1 system, for k-eff estimation during normal storage condition as well as in a potential waterdrop accident. TRITON and KENO-VI sequence from SCALE6.1 system was applied to determine the k-eff for the basket and pool, loaded with spent fuel elements, during the normal and the accidental considered conditions. All the spent fuel elements were irradiated in the reactor core, at 22 MW power, being discharged after the maximum burn up is reached. k-eff values shown a good agreement with previous calculations performed using MCNP5 code.

This study aims to verifying the nuclear criticality criteria in spent fuel repository. The repos... more This study aims to verifying the nuclear criticality criteria in spent fuel repository. The repository is an open pool type that is used temporary to store the spent fuel elements removed from the core of 22 MW power of open pool research reactor. The repository could receive up to 800 of spent fuel elements of material testing reactor (MTR) type. Arrangement configuration of spent fuel elements in the repository must verify the criticality criteria during the storage that mean that the multiplication factor should not exceed 0.85. Two models were developed in this study using MCNP5 code to simulate the spent fuel elements configuration including; the basket model, and the storage pool model (contains the maximum permissible capacity). The result shows that the multiplication factors would be less than the permissible limit for the criticality criteria of the storage.

A modeling was performed in this study for iodine isotopes discharge from a hypothetical nuclear ... more A modeling was performed in this study for iodine isotopes discharge from a hypothetical nuclear research reactor of thermal power of 20 MW, due to hypothetical severe nuclear accident in case of loss of ventilation system. The committed effective doses CED for the public around the reactor site was calculated for various atmospheric stability classes, Pasquill categories (A-F), using health physics HOTSPOT 2.06 code. The model was applied for three cases; without explosion of water, with explosion of water, and fusion in air. The results in the first case show that the receptor received CED lower than the permissible dose at any downwind distance from the reactor. In the second case, the receptor located within the distances 1.2 Km to 4.5 Km from the reactor received CED slightly more than the permissible dose. And for the third case, the receptor located within 50 Km from the reactor received CED more than the permissible dose. All the previous results of CED belong to the reactor site stability class (F).

Kerntechnik, 2009

One of the most versatile types of heat exchangers used is the plate heat exchanger. It has princ... more One of the most versatile types of heat exchangers used is the plate heat exchanger. It has principal advantages over other heat exchangers in that plates can be added and/or removed easily in order to change the area available for heat transfer and therefore its overall performance. The cooling systems of Egypt's second research reactor (ETRR 2) use this type of heat exchanger for cooling purposes in its primary core cooling and pool cooling systems. In addition to the change in the number of heat exchanger cooling channels, the effect of changing the amount of mass flow rate on the heat exchanger performance is an important issues in this study. The inertia flywheel mounted on the primary core cooling system pump with the plate heat exchanger plays an important role in the case of loss of flow transients. The PARET code is used to simulate the effect of loss of flow transients on the reactor core. Hence, the core outlet temperature with the pump-flywheel flow coast down is fed...

Transporting and processing of radioisotopes and irradiated targets inside hot cells generate a s... more Transporting and processing of radioisotopes and irradiated targets inside hot cells generate a significant contamination. The majority of contamination comes from dispersion of radioactive materials during processing the samples after irradiation. Processing includes opening, extracting the irradiated samples, and preparing the samples in a shield prior to transportation. A model of dispersion of radioactive products inside the cell is postulated. Before decontaminating the cell, the expected dose received by the worker must be evaluated. A RESRAD-BUILD code is used in this study to calculate the dose and the corresponding risk. The calculated dose received during the decontamination process is more than the permissible dose and many proposals are presented in the study to decrease the level of received doses.

Academia Letters, 2021

A new technique was introduced in this study to utilize the prompt gamma rays, that generated fro... more A new technique was introduced in this study to utilize the prompt gamma rays, that generated from the fission process, in gamma irradiation experiments. The method de- pends on utilizing the difference of the attenuation effect of water on neutrons from that of gamma rays
The core of 22 MW power is submerged in the water of open pool type reactor and the theoretical and experimental studies have illustrated that the influence of water in attenuating the prompt neutrons is higher than the prompt gamma rays.
Based on this phenomena, the dose rate distributions for the prompt neutrons and the prompt gamma rays were studied in the water volume that surrounding the core during reactor operation. Then, the proper position, that achieves the considerable gamma dose rate as well as the minimum neutron dose rate, could be determined. An irradiation device was developed to contain the samples during the gamma irradiation process. The irradiation device would be located in the proper position that was chosen depending on the previous conditions. MAVRIC code was used to determine the dose rate distributions in the water surrounding the reactor core.

Academia Letters, 2021

This study suggests an alternative path to transport the cobalt device after irradiation in open-... more This study suggests an alternative path to transport the cobalt device after irradiation in open-pool-type reactor from the reactor core to the cobalt cell. The cobalt transport process in current path depends on using heavy shielded cask, 1500 Kg weight, under water surface of spent fuel storage pool which may cause a missile impact of spent fuel accident (Falling of heavy load into spent fuel storage pool). To avoid using the shielded cask, an alternative path is suggested depending on transporting the cobalt device through the hot cells of the reactor; the testing and the cobalt cells. The expected radiation dose rate rise associated with the alternative transport path must be evaluated at positions which the worker will be found during transport process. SCALE/MAVRIC sequence code was used to determine the dose rate levels along the suggested transport path; over spent fuel storage pool water surface, and around testing and cobalt cells. The calculated results show rise in radiation dose levels more than the permissible dose rate limits around testing cell. Suggested procedures are presented in this study to decrease the received dose by worker found around testing cell during transport process.

KERNTECHNIK, 2010

During the irradiation process of MoO3 powder with in a high neutron flux, energy deposited in th... more During the irradiation process of MoO3 powder with in a high
neutron flux, energy deposited in the powder must be released
to avoid energy accumulation. The temperature increasing in
the powder due to energy accumulation may cause powder
phase change and under certain conditions the temperature
may reach the melting point. An analytical model was developed
to study the effect of gap conductance on the temperature
distribution in the powder during the irradiation process. The
gap conductance model was studied for helium and nitrogen
gases at different pressures to obtain the optimum conditions
verifying the safety in heat transfer point of view. It was found
that the gap conductance is a function of gas pressure. The
model provided the optimum conductivity for the gap filled
with helium gas at 1 atm pressure.

KERNTECHNIK, 2009

A one dimensional steady state thermal analytical model has been developed to study the heat tran... more A one dimensional steady state thermal analytical model has
been developed to study the heat transfer and temperature distributions
in a quartz ampoule filled with MoO3 powder. The
source heat generation inside the ampoule is released from the
high neutron flux (1.4 · 1014 neutrons/cm2s) interaction with
MoO3. Natural and forced convections heat transfer boundary
conditions are adopted during the irradiation process. The
peak temperatures in MoO3 powder and quartz are calculated
and compared with their melting temperatures to ensure the irradiation
safety criteria.

• This study calculated the dose rate in the control room due to air contamination. • The air con... more • This study calculated the dose rate in the control room due to air contamination. • The air contamination resulted from fission products release from a degraded core. • MCNP5 code was used to calculate the dose rate in the control room. • A suggestion to venting the contaminated air to the environment was studied. • The suggestion aims to decrease the dose rate in the control room. • The maximum dose resulting from the venting would locate in a free-inhabitant area. • The suggestion is adequate. A B S T R A C T The objective of this paper is to estimate radiation dose level in the control room due to air contamination in the containment of open pool type reactor during emergency situation. A postulated core degradation accident, due to fuel element blockage, causes fission products to release to the reactor containment and then, the ventilation system would convert automatically from the normal situation to the emergency situation to purify the contaminated air by forcing it to pass through a group of filters. The study computed the dose rate in the control room, adjacent to the containment, during the emergency situation. The dose rate level in the control room depends on the degree of the core degradation and so, the maximum dose rate is corresponding to the complete degradation of the core and the minimum dose rate is corresponding to the degradation of one fuel plate. The dose rate in the control room was calculated using MCNP5 code and it ranged from 1.12 to 2.03E-3 Sv/ h. The results also show that the dose rate level in the control room would continue having values more than the permissible limit for a long time and so, a suggestion of venting the radioactive isotopes from the containment into the environment was studied to decrease the dose rate level in the control room. The suggestion was possible because the maximum dose, resulting from the contaminated air venting, would fortunately locate in a free-inhabitant area.

KERNTECHNIK, 2019

As a perspective plan to transport the spent fuel of material testing of research reactor (MTR) f... more As a perspective plan to transport the spent fuel of material testing of research reactor (MTR) from the temporary storage to permanent storage, choosing an adequate cask is very important to ensure the safety precautions during the transport process. Latin America cask is one of the transportation cask types may be chosen to transport the spent fuel elements where it was designed to transport the irradiated fuel for MTR and the TRIGA research reactors. Therefore, it must be evaluated from the neutronic, radiological, and thermal points of view. The cask has internal diameter of 60 cm which make it possible to content 21 of spent fuel elements of MTR. This study aims to evaluate dose rate distribution around the cask after loading with 21 of MTR spent fuel elements which have been stored for 5-years as a minimal decay time in the temporary storage. For this, MCNP5 code was used to determine the dose rate in the radial and axial directions around the cask. The results show that the dose rates at the cask surface and at 200 cm from the surface are lower than the permissible transportation limits. Radiologische Bewertung von Transportbehältern für abge-brannte Brennelemente. Für den Transport des abgebrannten Brennstoffs des Materialprüf-Forschungsreaktors (MTR) vom Zwischenlager ins Endlager ist die Wahl eines geeigneten Be-hälters sehr wichtig, um die Sicherheitsvorkehrungen während des Transportvorgangs zu gewährleisten. Der lateinamerikani-sche Behälter ist einer der Transportbehältertypen, die für den Transport der abgebrannten Brennelemente gewählt werden können, da er für den Transport des bestrahlten Brennstoffs für MTR und die TRIGA-Forschungsreaktoren konzipiert wurde. Dies muss unter neutronischen, radiologischen und thermischen Gesichtspunkten bewertet werden. Der Behälter hat einen Innendurchmesser von 60 cm, was es ermöglicht, 21 abgebrannte Brennelemente der MTR zu enthalten. Ziel dieser Studie ist es, die Dosisleistungsverteilung um den Behäl-ter herum nach der Beladung mit 21 abgebrannten MTR-Brennelementen, die 5 Jahre lang als minimale Zerfallszeit im Zwischenlager gelagert wurden, zu bewerten. Dazu wurde mit dem MCNP5-Code die Dosisleistung in radialer und axialer Richtung um den Behälter herum bestimmt. Die Ergebnisse zeigen, dass die Dosisleistungen an der Behälteroberfläche und bei 200 cm von der Oberfläche niedriger sind als die zuläs-sigen Transportgrenzen.

applied radiation and isotopes, 2017

The aim of this study was to evaluate skin dose received by a worker contaminated accidently with... more The aim of this study was to evaluate skin dose received by a worker contaminated accidently with radiochemical compounds used in Mo-99 extraction process. The skin contamination calculations are executed for four stages of chemical Mo-99 extraction process using VARSKIN code. The worker was received a very high radiation skin contamination values because the radiochemical compounds contain dissolved fission products with very high specific activities. The expected time to decontaminate the skin is also calculated. The results show that the delay in decontamination process will expose the worker rapidly to radiation dose values reach to the threshold of onset of skin injury.

This study aims to calculate the dose rate profiles after cobalt device ejection from open-pool-t... more This study aims to calculate the dose rate profiles after cobalt device ejection from open-pool-type reactor core. MicroShield code was used to evaluate the dose rates inside the reactor control room. McSKY code was used to evaluate the dose rates outside the reactor building. The calculated dose rates for workers are higher than the permissible limits after 18 s from device ejection. a b s t r a c t The evaluation of the radiation dose during accident in a nuclear reactor is of great concern from the viewpoint of safety. One of important accident must be analyzed and may be occurred in open pool type reactor is the rejection of cobalt device. The study is evaluating the dose rate levels resulting from upset withdrawal of co device especially the radiation dose received by the operator in the control room. Study of indirect radiation exposure to the environment due to skyshine effect is also taken into consideration in order to evaluate the radiation dose levels around the reactor during the ejection trip. Microshield, SHLDUTIL, and MCSky codes were used in this study to calculate the radiation dose profiles during cobalt device ejection trip inside and outside the reactor building.

SSRN, 2023

Nuclear science and engineering, 2023

KERNTECHNIK, 2022

ASME Journal of nuclear engineering and radiation science, 2021

Ssrn electronic journal , 2022

BRAZILIAN JOURNAL OF RADIATION SCIENCES, 2022

JOURNAL OF NUCLEAR RESEARCH AND DEVELOPMENT, NO. 22, DECEMBER 2021, 2021

Kerntechnik, 2009

Academia Letters, 2021

KERNTECHNIK, 2010

KERNTECHNIK, 2009

KERNTECHNIK, 2019

applied radiation and isotopes, 2017

Ssrn electronic journal, 2022