Расчетно-экспериментальные исследования радиационно-защитных свойств композитов на основе природных минералов Республики Ирак тема диссертации и автореферата по ВАК РФ 00.00.00, кандидат наук Альсафи Ханин Махмуд Джабер

• Альсафи Ханин Махмуд Джабер
• кандидат наук
• 2026

Оглавление диссертации кандидат наук Альсафи Ханин Махмуд Джабер

Content

CHAPTER 1. Review and analysis of the state of radiation safety of nuclear facilities and entities of the Republic of Iraq

1.1. Overview of Nuclear Power Facilities in the Republic of Iraq

1.1.1. Early development of nuclear science in Iraq

1.1.2. Establishment of the Tuwaitha Nuclear Research Centre

1.1.3. Expansion of nuclear facilities (1970s-1980s)

1.1.4. Destruction of Iraqi nuclear facilities

1.2. Review and analysis of sources of radioactive contamination and waste

1.2.1. Isotopic composition of radioactive contaminants generated during operation

1.2.2. Environmental pollution with man-made radionuclides

1.3. Methods of localization and isolation of radioactive waste

1.3.1. Environmental issues of radioactive contamination of nuclear facilities and territories of the Republic of Iraq

1.3.2. Review of options for radioactive waste containment and isolation in the Republic of Iraq

1.3.3. Review of studies of radiation-protective properties of local natural minerals and composites made from them

1.3.3.1. Epoxy resin-based composites

1.3.3.2. Clay matrix-based materials

1.4. Modern materials for radiation protection, developed at the Ural Federal University using natural raw materials

1.4.1. Development of protective materials based on a cement matrix

1.4.2. Clay bricks for radiation protection

1.4.3. Ceramic shielding materials based on natural minerals

1.5. Environmental and engineering advantages of rare earth metals based on natural

minerals

1.6. Conclusions to Chapter

CHAPTER 2. MATERIALS, EQUIPMENT AND METHODSRESEARCH

2.1. Description of materials

2.1.1. WhiteIraqi sand

2.1.2. Iraqi clay (Clay 1 and Clay 2)

2.1.3. Lead oxide (PbO) and titanium dioxide (TiO2)

2.1.4. Epoxy resin and hardener

2.1.5. Boric acid (H3BO3)

2.1.6. Glass waste powder

2.1.7. Granite powder

2.2. Sample preparation

2.2.1. Glass samples

2.2.2. Manufacturing of epoxy resin-based composites

2.2.3. Calcinated clay composites

2.2.3.1. The difference between calcinated clay and ceramics samples

2.2.3.2. Fabrication of the calcinated Iraqi clay-based composites

2.3. Experimental methods for determining the characteristics of the samples under study

2.3.1. X-ray diffraction

2.3.2. Fourier transform infrared spectroscopy

2.3.3. Energy-dispersive X-ray spectroscopy

2.3.4. Physical properties

2.3.5. Experimental measurements of gamma radiation protection parameters

2.4. Calculation methods for determining the characteristics of samples

2.4.1. Monte Carlo N-Particle Transport Code (MCNP-5)

2.4.2. Calculation using the XCOM program

2.5. Determination of statistical error for experimental measurements

2.6. Reliability of results

2.7. Conclusions to Chapter

CHAPTER 3: Study of radiation protection characteristics

3.1. Protective glass

3.1.1. Shielding properties of silicate glasses based on titanium dioxide

3.1.2. Shielding properties of silicate glasses based on lead oxide

3.1.3. Mechanical properties of silicate glasses based on lead oxide and titanium dioxide

3.1.4. Comparison of manufacturing costs and shielding efficiency of the developed glasses with other samples

3.2. Epoxy resin composites

3.2.1. Density and effective atomic number (Zeff) measurements

3.2.2. X-ray diffraction

3.2.3. Scanning electron microscopy analysis

3.2.4. Gamma radiation protection efficiency of developed epoxy resin-based clay composites

3.3. Composites based on calcined clay

3.3.1. Composite calcined clay with boric acid

3.3.1.1. X-ray diffraction

3.3.1.2. Infrared spectroscopy analysis

3.3.1.3. Scanning electron microscopy analysis

3.3.1.4. Properties of developed composites based on boric acid for protection against gamma radiation

3.3.2. Fired clay (clay 1 and clay 2) mixed with glass waste powder

3.3.2.1. X-ray diffraction

3.3.2.2. Infrared spectroscopy and analysis of samples calcined using glass waste (clay 1 and clay 2)

3.3.2.3. Physical and chemical properties of CBGW and SBGW composites

3.3.2.4. Radiation shielding properties of CBGW and SBGW composites

3.3.3. Fired clay (Clay 1 and Clay 2) mixed with granite (CBG and SBG)

3.3.3.1. X-ray diffraction

3.3.3.2. Physical and chemical properties

3.3.3.3. Radiation protection properties of CBG and SBG-based composites

3.4. Application areas of the developed composites

3.5. Conclusions to Chapter

CONCLUSION

List of abbreviations and symbols Bibliography

163

Введение диссертации (часть автореферата) на тему «Расчетно-экспериментальные исследования радиационно-защитных свойств композитов на основе природных минералов Республики Ирак»

Introduction

Relevance of the research topic and the degree of its development. Throughout the history of human development of atomic energy and other radiation technologies, radiation safety has always been a challenging area. As statistical data on the effects of radiation on humans accumulated, specialists developed new approaches to minimizing radiation doses, and radiation safety standards were revised and tightened. Numerous works by scientists around the world have been devoted to radiation safety issues. Therefore, it is impossible to single out any individual. Even a brief list of outstanding scientists who have made significant contributions to radiation safety would take several pages of introduction.

However, despite improvements in the safety of new and existing nuclear power plants (NPPs), including in the area of radiation safety, the number of new NPP units is increasing, integrated radiation doses are increasing as the operating life of existing NPPs increases, and large-scale decommissioning of NPPs at the end of their service life is underway. The ever-expanding use of radiation technologies in various fields of human activity, particularly in medicine, also requires the adoption of effective measures to ensure radiation safety.

One of the main principles of radiation safety is standardization.

The annual dose limit established for personnel in 1956 (50 mSv) [1], did not change until 1990, then in accordance with the recommendations International Commission on Radiological Protection (ICRP) (Publication 60) [2], was reduced to 20 mSv per year (with the possibility of averaging) based on a revision of the risk estimates of stochastic effects obtained from long-term studies of radiation effects in survivors of the atomic bombings of Hiroshima and Nagasaki. The recognition of the absence of a safe level of radiation led to the idea of reducing exposure as much as possible. However, radiation protection, like other practical tasks, is subject to the "law of diminishing returns". Therefore, optimization of

radiation protection is necessary. The absence of an observed threshold dose and the limited availability of resources motivated the development of the ALARA (As Low As Reasonably Achievable) principle, taking into account economic and social factors. Finland, the first country in the world, adopted the recommendations of Publication 60 into legislation (effective 1 January 1992).

The radiation dose is proportional to the value of the radiation parameter (e.g., the radiation dose rate) and the time spent in the radiation field, and is inversely proportional to the square of the distance from the radiation source to the worker. Therefore, the radiation dose can be reduced by shortening the duration of work in conditions of ionizing radiation (e.g., using automated equipment or conducting training in a clean area), increasing the distance between the radiation source and the worker (e.g., using remote-controlled devices), and reducing the value of the radiation parameter [3].

Among the methods for reducing radiation dose, the most common is shielding with radiation-protective material.

The continued expansion of nuclear energy and radiation technologies is driving industry and research to develop new radiation-protective materials (REM) with good protective properties and low toxicity, including composites that allow their composition to be tailored to the intended irradiation conditions. Despite the large number of developments, the search for REMs that are high-tech in manufacturing and easy to use. An important condition is the selection (optimization of the composition) of REEs applicable to the planned irradiation conditions in accordance with the requirements of ICRP Publication 1031.

Expansion of the use of radiation technologies in various areas of activity initiated a rapid growth of scientists' attention to development of new radiation-protective materials (REM) with good protective properties and low toxicity. According to Scopus data, 13,031

1ICRP, 2007. The 2007 Recommendations of the International Commission on Radiological Protection. ICRP Publication 103. Ann. ICRP 37 (2-4).

articles in the field of radiation protection materials were published between 2020 and February 2026 (Figure B1). Of these, the largest number were devoted to protective glass (2,588 articles), followed by polymers and polymer composites for radiation protection (1,606 articles). Ceramic-based protective materials were the subject of 498 articles, cement matrix materials - 1,107 articles, and clay (protective bricks) - 88 articles.

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Figure B1: Number of published articles devoted to various radiation-protective materials (ceramics, polymers, glass, concrete, brick, alloys)

However, almost all studies are not related to the study of local mineral deposits for use in the construction of nuclear facilities in a particular country. Therefore, taking into account the recommendations ICRP Publication 103 to strengthen the implementation of the optimization principle, it is especially important to highlight the relevance b studies of radiation-protective properties of local natural minerals, as well as composite radiation-protective materials developed on the basis of these minerals, including those incorporating

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industrial waste, to assess the potential for their use in radiation protection of facilities in countries where nuclear energy facilities (NEFs) are being constructed.

Such complex research has been conducted at UrFU for about 20 years under the scientific supervision of Professor, Doctor of Technical Sciences O.L. Tashlykov. During this period, computational and experimental studies have been conducted on various radiation-protective composite materials in collaboration with scientists from many countries: Dr. Mohammad Ibrahium Abu-Alsayyed (Jordan), Mohammad Wasef Marashdeh and Mohammed Hassan Abu-Mhareb (Saudi Arabia), Islam Nabil, Fawzy Hammad Sallem (Republic of Egypt), Nasser KA (India), Kawa Kaky (Iraq), Daria Tishkevich (Belarus).

At the request of their respective countries' ministries, graduates from universities in these countries are sent to UrFU for postgraduate studies to conduct research and potential assessments of the potential use of local natural minerals and composites made from them in the field of radiation protection.

Karem Mahmoud Abdelazim Gaber (Republic of Egypt) successfully completed postgraduate studies, completed the required amount of research, and defended their dissertations for the degree of Candidate of Technical Sciences [4], Aladailah Mutaz Walid Ali (Jordan) [5], Ta Van Thuong (Vietnam) [6]. As studies show, the use of local natural raw materials in the construction of nuclear facilities, including radioactive waste storage facilities allows to reduce construction costs by an amount up to 30%.

For the Republic of Iraq, the relevance of research in this area is significantly increasing due to the need to carry out a huge amount of work on the reliable localization of radioactive waste and significant radioactive contamination of the country's territory formed As a result of military action using depleted uranium munitions, as well as the destruction of nuclear research reactors due to the country's limited financial resources. To a certain extent (although not specified in international regulations), decommissioning of the destroyed reactors may also be considered.

The most challenging issue in practice is protection against photon radiation, as protection against alpha and beta radiation is not a problem, while exposure to neutron radiation is possible only during the operation of a nuclear reactor or in very rare cases in the presence of neutron radiation sources. Heavy materials (lead, tungsten, depleted uranium, etc.) are the most effective in protecting against gamma radiation. Lead is the most commonly used of these. However, it has a number of disadvantages; in particular, its high plasticity requires special designs and fastenings when installing vertical shielding made of lead thicker than 5 mm [7]. In addition, lead is a toxic substance, which limits its use and determines additional protective measures when disposing of lead-containing radiation-protective materials (REM).

Ceramic-based materials are among the most promising candidates due to their exceptional thermal stability, mechanical strength, corrosion resistance, and ability to incorporate various heavy or neutron-absorbing oxides [8].

In recent years, research efforts have increasingly focused on developing new non-toxic compounds, including lead-free rare earth (REE) composite materials, with high protective properties. Radiation shielding materials must meet a number of requirements to ensure long-term effectiveness and maximum protection. These requirements include structural strength, radiation and heat resistance, chemical inertness, high thermal conductivity, low coefficient of linear expansion, affordability, and abundance. Many of these criteria are mutually exclusive, and no single material satisfies all requirements. However, by optimizing the composition, it is possible to identify materials or combinations of materials that largely meet these requirements. A number of effective shielding materials have been developed and studied [9]. The cost of radiation protection for modern nuclear power plants can account for 20-30% of the total construction costs [3], therefore, even a small reduction in the cost of materials for the construction of nuclear facilities leads to a significant reduction in the costs of constructing such facilities.

In this regard, the development of new materials (glass, coatings, fired clay, and others) with high protective capacity against gamma radiation is a pressing task for the creation of new technologies and equipment models that advance the priority area of science, technology, and engineering in the Russian Federation: paragraph 8 (Energy efficiency, energy conservation, and nuclear energy), as well as in other countries around the world, especially in areas where the Rosatom State Corporation is present.

The aim of the dissertation work is the development of new, effective and affordable radiation-protective materials based on natural minerals from the Republic of Iraq for use in the construction of biological protection for nuclear facilities, including radioactive waste storage facilities, and other radiation-hazardous facilities.

The subject of the study is the radiation-protective properties of natural minerals of the Republic of Iraq and composite radiation-protective materials made on their basis.

To achieve this goal, the following tasks were set and solved:

1. Review and analysis of the state of radiation safety of nuclear facilities and entities of the Republic of Iraq.

2. Selection of natural resource deposits most promising for the research area under consideration, collection of minerals and their transportation from Iraq to UrFU.

3. Conducting comprehensive computational and experimental studies of the radiation-protective characteristics of natural minerals of the Republic of Iraq to assess their potential use in radiation protection of nuclear facilities, including when handling radioactive waste.

4. Development of the composition, production technology, sample preparation and experimental studies of the radiation-protective properties of glass made from Iraqi sand

5. Development of the composition, production technology, sample manufacture, and conducting computational and experimental studies of the radiation-protective properties of composite REEs based on an epoxy resin matrix with a clay filler.

6. Development of the composition, production technology, sample manufacture, and conducting computational and experimental studies of the radiation-protective properties of composite REEs based on a clay matrix, including those with industrial waste as a filler.

7. Conducting computational and experimental studies of the influence of pressure and temperature during the manufacturing process on the radiation-protective properties of ceramic samples.

Scientific novelty of the work consists of the following:

1. For the first time, comprehensive computational and experimental studies of the radiation-protective characteristics of natural minerals of the Republic of Iraq were conducted to assess the potential for their use in radiation protection of nuclear facilities, including in the handling of radioactive waste.

2. New glass formulations, including lead-free ones, have been developed using sand from deposits in the Republic of Iraq (Anbar Province), samples have been manufactured, and computational and experimental studies of their radiation-protective properties have been conducted.

3. For the first time, compositions were developed, samples were manufactured, and computational and experimental studies were conducted on the radiation-protective properties of composite REEs based on a polymer matrix (epoxy resin) with fillers in the form of natural minerals of the Republic of Iraq.

4. For the first time, compositions were developed, samples were manufactured, and calculation and experimental studies were conducted on the radiation-protective properties of composite rare earth metals made from two types of clay from deposits in the Republic of Iraq, including with the addition of industrial waste (glass and granite).

Theoretical and practical significance of the work:

• Results of studies of radiation-protective properties of natural minerals of the Republic of Iraq will be used to evaluate their potential use in the construction of radiation

protection for nuclear facilities, radioactive waste storage facilities, and the aftermath of the destruction of the nuclear reactors at the Tuwaitha Research Centre.

• The results of calculation and experimental studies of the dependence of radiation-protective properties of samples on pressure and temperature can be used in organizing the industrial production of ceramic protective blocks.

• The results of computational and experimental studies of the effect of heavy metal additives in a glass matrix based on Iraqi sand on radiation-protective properties can be used in the development and production of glass based on sand from the Republic of Iraq.

• The developed and tested methodology of the conducted computational and experimental studies will be used as a basis for conducting similar studies of other minerals and composite radiation-protective materials based on them.

• Positive results of the assessment of the influence of various fillers on the radiation-protective properties of composites made from local natural minerals of the Republic of Iraq make it possible to partially solve the problem of industrial waste disposal.

Methodology and methods of dissertation research.

The dissertation used modern equipment and experimental research methods: Scanning electron microscopy (SEM) was used to study the microstructure and composition of the samples; the chemical composition was analyzed using energy-dispersive X-ray analysis (EDXA); the crystal structures of the samples were determined by X-ray diffraction; Fourier transform infrared spectroscopy made it possible to identify functional groups and intermolecular chemical bonds, as well as to observe changes in the material. The results of experimental measurements of the protective characteristics of the samples were confirmed by theoretical calculations using XCOM software based on the National Institute of Standards and Technology (NIST) nuclear library database, as well as by Monte Carlo simulations using the MCNP-5 code and the ENDF/B-VI nuclear library database.

Provisions submitted for defense:

• The relevance of using local natural minerals and composite materials made from them for the construction of radiation protection for nuclear facilities, radioactive waste storage facilities, and the elimination of the consequences of radioactive contamination of the territories of the Republic of Iraq.

• The results of the dissertation determine the potential for using local natural minerals in the construction of radiation protection for nuclear facilities and radioactive waste storage facilities in the Republic of Iraq.

• Composite REEs based on a clay matrix using granite and glass powder as filler have significant potential to reduce the cost of biological protection during the construction of radioactive waste storage facilities in the Republic of Iraq.

• Composite REEs based on an epoxy resin matrix with clay filler make it possible to create a protective coating with high protective and waterproofing properties for the supporting structure of radioactive waste storage facilities.

• Using sand from deposits in the Republic of Iraq allows for the production of glass, including lead-free glass, with high radiation-protective properties and a lower cost than foreign analogues.

The degree of reliability of the results obtained is based on a comprehensive analysis of previously completed work on the subject of research, the use of modern research tools, modeling and calculation methods, proven software, verified instruments and measuring systems, good convergence and the results obtained experimentally, with the results of modeling using the MCNP-5 calculation code, the XCOM program, as well as with the results obtained by other authors.

Personal contribution of the author consists of in the selection and justification of research areas; selection of deposits, collection, transportation of natural Iraqi minerals to Russia (to the Department of Nuclear Power Plants and Renewable Energy Sources of UrFU), their preparation for research; preparation of samples of radiation-protective materials (REM), development of experimental methods; direct participation in scientific

experiments; development of facilities; mathematical processing of experimental data; computer modeling using the Monte Carlo method; preparation of publications, reports at conferences, scientific and technical seminars; discussion of the results of work with organizations involved in the creation/implementation of the developed technologies. All presented materials were obtained by the author personally or in collaboration.

Testing the results of the work

The main results of the dissertation were discussed and received approval.at 9 scientific conferences, including:

1. 15th International Symposium on Radiation Physics "ISRP-15", Kuala Lumpur, Malaysia, December 6-10, 2021

2. International Conference on Applications of Neutron Scattering in Condensed Matter Research "RNICS-2021", Yekaterinburg; September 27 - October 1, 2021

3. XXXIII Russian Youth Scientific Conference with International Participation "Problems of Theoretical and Experimental Chemistry", Yekaterinburg, 2023.

4. X International Youth Scientific Conference on Physics, Technology and Innovation "PTI-2023", Yekaterinburg, May 15-19, 2023

5. International Conference on Sustainable Technologies in Civil and Environmental Engineering "ICSTCE 22023", Pune, India, 15-16 June 2023

6. IX Baikal International Conference "Magnetic Materials. New Technologies" (BIKMM-2023), Baikalsk, September 11-14, 2023

7. Conference on the Application of Neutron Scattering in Condensed Matter Research "RNICS-2023", Yekaterinburg, September 25-28, 2023

8. XI International Youth Scientific Conference "Physics. Technologies. Innovations." FTI-2024, Yekaterinburg, May 20-25, 2024.

9. XXI International Scientific and Practical Conference "Nuclear Energy Safety", Volgodonsk, October 23-24, 2025.

Publications: On the topic of the dissertation, 23 scientific papers have been published, including 8 published in peer-reviewed scientific journals determined by the Higher Attestation Commission of the Russian Federation and the Ural Federal University Attestation Council, and indexed in the international citation databases WoS and Scopus.

Structure and scope of work. The structure of the dissertation research is subordinated to the concept of the research and consists of an introduction, three chapters, and a conclusion a list of references, including 146 titles. Total volume of the dissertation 181 pages. The work contains 55 drawings and 19 tables.

Заключение диссертации по теме «Другие cпециальности», Альсафи Ханин Махмуд Джабер

CONCLUSION

The results presented in this dissertation collectively highlight significant progress in the development of radiation shielding materials from locally available and affordable Iraqi raw materials. The overall goal of this research was to create cost-effective materials that could serve as alternatives in various radiation shielding applications.

1- A detailed analysis of radioactivity distribution was conducted in various regions of Iraq, including the area adjacent to the destroyed research reactor (IRT 5000, Tamuz-2). Radioactivity analysis shows high concentrations of Cs-137 and Co-60, reaching 14,772.41 ± 99.91 Bq/kg and 7,642.22 ± 40.02 Bq/kg, respectively. Furthermore, due to the war, some regions were contaminated with depleted uranium.

2- Three types of Iraqi raw materials were collected and donated to UrFU University: sand from the city of Anbar and two clay minerals, both widely and inexpensively mined in various regions of Iraq (clay 1 from the city of Babylon) and clay 2 from the city of Diwaniya).

3- Based on the collected sand, two series of glasses were produced, the first series containing different concentrations of TiO2, while the second contained approximately 35 wt.% PbO.

4- In addition, based on the collected samples of clay 1 and clay 2, two polymer series were produced, as well as five series of fired clay in the form of a mixture of clay 1 and clay 2 with boric acid, glass production waste and granite.

5- Experimental measurements and theoretical modeling of the radiation-protective properties of the manufactured composites (glass, polymer, fired clay) were carried out.

6- At a gamma energy of 0.662 MeV, the cost of manufacturing glass based on Iraqi sand with 35 wt.% PbO is approximately 30% lower than the cost of commercial protective glass RS 323 G19. At the specified energy, the half-value layer of both the manufactured glass with 35 wt.% PbO and the commercial protective glass RS 323G19 is equivalent to 2.48 cm.

7- At a gamma radiation energy of 0.662 MeV, the glasses produced with the addition of approximately 1-12 wt.% TiO2 glasses have half-value layer as thick as those of commercial protective glasses RS 253 and RS 253 G18, by 3-7%. In addition, the cost of manufacturing glasses based on Iraqi sand and TiO2 compounds is 25-31% lower than that of protective glasses RS 253 and RS 253 G18.

8- At the gamma energy of 0.662 MeV, the addition of clay 1 to epoxy resin-based composites decreases their half-value layers by ~ 28% when the concentration of clay 1 increases to 60 wt.%, while the addition of clay 2 to epoxy composites decreases their half-value layers by 21% when the concentration of clay 2 increases to 60 wt%. In addition, the fabrication costs of epoxy composites based on clay 1 or clay 2 were reduced.

9- At the energy of 0.662 MeV, the half-value layer of the calcinated clay 1 with B2O3 additives increases by ~ 11% when the B2O3 concentration increases to 20 wt%. In addition, the fabrication cost of the calcinated clay 1 with B2O3 additives increases by 98-294% when the B2O3 concentration increases to 20 wt%.

10- At 0.662 MeV, there is no significant improvement in the half-value layer of the composite made from calcinated clay 1 or clay 2 with added waste glass. On the other hand, the fabrication cost of calcinated clay 1 and clay 2 with added waste glass increases by 29% as the waste glass concentration increases to 50 wt%.

11- At an energy of 0.662 MeV, the half-value layer of calcinated clay 1 with the addition of granite decreases by ~ 18% with an increase in the granite concentration to 50 wt%.

On the other hand, the fabrication cost of calcinated clay 1 with the addition of granite increases by 5% when the granite concentration increases to 50 wt%.

12- At an energy of 0.662 MeV, the half value-layer of calcinated clay 2 with the addition of granite decreases by ~ 32% with an increase in the granite concentration to 50 wt.%. On the other hand, the fabrication cost of calcinated clay 2 with the addition of granite increases by 1% with an increase in the granite concentration to 50 wt.%.

13- The average equivalent dose rate at a distance of 1 m from the protected radioactive waste is maximum and reaches 1085 ^Sv/h. After this, it is sufficient to use a container with a wall thickness of 50 cm (40 cm of CBG50 composite (clay 1 + 50 wt% granite) + 10 cm of E-C60 composite (epoxy resin + 60 wt% clay1)) to reduce the effective dose rate from buried radioactive waste to 1.85 ^Sv/h (permissible levels).

14- In addition, the study also shows that the net thickness of the radioactive waste container of 50 cm (40 cm of SBG50 composite (clay 2 + 50 wt% granite) + 10 cm of E-S60 composite (epoxy resin + 60 wt% clay 2)) is sufficient to reduce the equivalent dose rate from buried radioactive waste to 1.47 ^Sv/h (permissible levels).

Prospects for further development of this research topic include the development of the

following areas:

1. Instead of using calcinated clay as a simple aggregate, utilize it as the primary precursor for a geopolymer matrix. Incorporate nano-sized boric acid and nano-granite powder. This approach maximizes the surface area for neutron capture (via boron) and enhances the packing fraction, leading to superior mechanical strength and a more homogeneous attenuation of both gamma and neutron radiation compared to ordinary Portland cement (OPC)-based composites.

2. Utilizing Iraqi calcined clay in the development of geopolymer cements designed for the immobilization (encapsulation) of radioactive waste.

3. Develop a printable geopolymer ink based on Iraqi calcinated clay, finely ground glass waste, and boric acid. Use 3D printing to create shields with complex internal

geometries—such as honeycombs or gyroid lattices—that are impossible to achieve with traditional casting. These structures can be optimized for minimal weight while maximizing path length (scattering) for radiation, allowing for the on-site, automated construction of custom-shaped shielding blocks for nuclear facilities.

4. Iraq's climate involves high temperatures and potential humidity. Investigate the durability of the composite under hydrothermal conditions. A novel approach would be to pre-treat the calcinated clay and granite with silane coupling agents before mixing with boric acid and glass waste. This creates a hydrophobic interface within the matrix, preventing the leaching of boric acid (which is water-soluble) and maintaining the material's neutron attenuation coefficient over decades of service in humid environments.

5. Introduce controlled porosity using aerating agents to create lightweight panels. While porosity typically degrades shielding, a novel approach is to engineer the pore walls to be dense with boric acid and granite nanoparticles, and to backfill the pores with a high-density epoxy or paraffin wax infused with additional neutron absorbers. This yields a "sandwich" structure at the micro-scale that is lightweight, structurally sound, and radiological resistant.

6. Introduce a conductive phase (such as carbon nanotubes or magnetite, which is often present in Iraqi granite) to the calcinated clay matrix. Evaluate the dual-functionality of the composite: shielding against ionizing radiation (gamma/neutron) while simultaneously providing electromagnetic interference shielding. This is critical for protecting sensitive electronics in nuclear power plants or medical diagnostic labs.

7. Leverage the pozzolanic reactivity of Iraqi calcinated clay combined with the alkali-silica reaction potential of glass waste. Engineer a mortar where micro-cracks induced by thermal stress (common in nuclear facilities) can self-heal through the dissolution of unreacted glass particles and secondary geopolymerization, ensuring long-term shielding integrity without manual intervention.

Список литературы диссертационного исследования кандидат наук Альсафи Ханин Махмуд Джабер, 2026 год

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