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        <full_title>EARTH SCIENCES AND HUMAN CONSTRUCTIONS</full_title>
        <issn media_type="print">2944-9154</issn>
        <issn media_type="electronic">2944-9006</issn>
      </journal_metadata>
      <journal_article>
        <titles>
          <title>Enhancing Structural Safety of Reactor Pressure Vessels through Probabilistic Fracture Mechanics Modeling</title>
        </titles>
        <contributors>
          <person_name sequence="first" contributor_role="author">
            <given_name>Md. Sifatul</given_name>
            <surname>Muktadir</surname>
            <affiliations>
              <institution>
                <institution_name>Department of Nuclear Engineering, Military Institute of Science and Technology, Dhaka, BANGLADESH</institution_name>
              </institution>
            </affiliations>
          </person_name>
          <person_name sequence="additional" contributor_role="author">
            <given_name>Udayan Das</given_name>
            <surname>Niloy</surname>
            <affiliations>
              <institution>
                <institution_name>Institute of Nuclear and Power Engineering, Bangladesh University of Engineering Technology, Dhaka, BANGLADESH</institution_name>
              </institution>
            </affiliations>
          </person_name>
          <person_name sequence="additional" contributor_role="author">
            <given_name>Abdus Sattar</given_name>
            <surname>Mollah</surname>
            <affiliations>
              <institution>
                <institution_name>Department of Nuclear Engineering, Military Institute of Science and Technology, Dhaka, BANGLADESH</institution_name>
              </institution>
            </affiliations>
          </person_name>
          <person_name sequence="additional" contributor_role="author">
            <given_name>Nazmul</given_name>
            <surname>Hassan</surname>
            <affiliations>
              <institution>
                <institution_name>Department of Nuclear Engineering, Military Institute of Science and Technology, Dhaka, BANGLADESH</institution_name>
              </institution>
            </affiliations>
          </person_name>
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          <jats:p>This study presents the development and application of a Probabilistic Fracture Mechanics (PFM) code to assess the failure probability of nuclear reactor pressure vessels (RPVs) with pre-existing cracks. RPVs, made of low alloy steel and subjected to various aging mechanisms such as fatigue and stress corrosion cracking, are especially vulnerable in the beltline region where nuclear fission occurs. A Python-based PFM code was developed to evaluate failure risk considering a single initial crack, whose size follows a log-normal distribution. The final (critical) crack size is calculated using applied design stresses and plane strain fracture toughness (KIC), which is modeled based on Nil Ductility Temperature (NDT) and neutron fluence, using IAEA and ASME recommendations. Stress input- membrane, bending, thermal, and seismic- are determined analytically using ANSYS simulation tools and Bangladesh National Building Code (BNBC) seismic guidelines. Failure probabilities are evaluated for both vertical and horizontal crack orientations, and results are benchmarked against PRAISE-JNES code outputs. The study finds that vertical cracks pose a higher failure risk, and that increased temperature and pressure significantly raise failure probability. The proposed PFM framework offers a reliable and compliant tool for probabilistic safety assessment of RPVs.</jats:p>
        </jats:abstract>
        <publication_date media_type="print">
          <month>12</month>
          <day>10</day>
          <year>2025</year>
        </publication_date>
        <publication_date media_type="online">
          <month>12</month>
          <day>10</day>
          <year>2025</year>
        </publication_date>
        <pages>
          <first_page>62</first_page>
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        <publisher_item>
          <item_number item_number_type="article_number">7</item_number>
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          <ai:license_ref>https://creativecommons.org/licenses/by/4.0/deed.en_US</ai:license_ref>
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          <doi>10.37394/232024.2025.5.7</doi>
          <resource>https://wseas.com/journals/eshc/2025/a14eshc-006(2025).pdf</resource>
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          <citation key="ref0">
            <unstructured_citation>I. A. E. A. (IAEA), Integrity of Reactor Pressure Vessels in Nuclear Power Plants, IAEA Nuclear Energy Series No. NP-T-3.11, 2009.</unstructured_citation>
          </citation>
          <citation key="ref1">
            <unstructured_citation>Rosatom, VVER-1200 Reactor Technical Overview., Rosatom, 2018.</unstructured_citation>
          </citation>
          <citation key="ref2">
            <unstructured_citation>U. N. R. C. (NRC), Reactor Pressure Vessel Integrity and Pressurized Thermal Shock, NUREG1806, 2007.</unstructured_citation>
          </citation>
          <citation key="ref3">
            <unstructured_citation>G. Was, Fundamentals of Radiation Materials Science: Metals and Alloys, Springers, 2007.</unstructured_citation>
          </citation>
          <citation key="ref4">
            <unstructured_citation>D. Blagoeva, Stability of Ferritic Steel to Higher Doses: Survey of Reactor Pressure Vessel Steel Data and Comparison with Candidate Materials for Future Nuclear Systems, International Journal of Pressure Vessels and Piping, 2, 2014.</unstructured_citation>
          </citation>
          <citation key="ref5">
            <unstructured_citation>G.R. Odette and G.E. Lucas, Embrittlement of nuclear reactor pressure vessels, JOM, vol. 53(7), 18-22, 2001.</unstructured_citation>
          </citation>
          <citation key="ref6">
            <unstructured_citation>II-Seok Jeong, Changheui Jang, Jun-Hyun Park, and Sung-Gyu Jeong, Lessons learned from the plant-specific pressurized thermal shock integrity analysis on an embrittled reactor pressure vessel, International Journal of Pressure Vessels and Piping, 78(2-3):99-109, 2001.</unstructured_citation>
          </citation>
          <citation key="ref7">
            <unstructured_citation>I. A. E. Agency, Pressurized Thermal Shock in Nuclear Power Plants, IAEA-TECDOC-1627, 2009.</unstructured_citation>
          </citation>
          <citation key="ref8">
            <unstructured_citation>Mehmet Ali Acar, Can Gönenli, and Mustafa Berkant Selek, Finite Element Analysis of the Pressure Vessels with Various Materials and Thicknesses, International Journal of Scientific Research and Management (IJSRM), Vol. 12(09), 1452-1459, 2024. DOI: 10.18535/ijsrm/v12i09.ec05/</unstructured_citation>
          </citation>
          <citation key="ref9">
            <unstructured_citation>Lih-jier Young, A fracture mechanics analysis of the PWR nuclear power plant reactor pressure vessel beltline weld, Journal of Nuclear Materials 288, 197-201, 2001.</unstructured_citation>
          </citation>
          <citation key="ref10">
            <unstructured_citation>Heba K. Louis, Afaf A.E. Ateya and Esmat Amin, Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel, Radiation Physics and Chemistry, 221(4),111738, 2024.DOI: 10.1016/j.radphyschem.2024.111738</unstructured_citation>
          </citation>
          <citation key="ref11">
            <unstructured_citation>R. K. Gupta, Probabilistic Fracture Mechanics Analysis of Reactor Pressure Vessel with Underclad and Through-Clad Cracks Under Pressurized Thermal Shock Transient, ResearchGate, 2018.</unstructured_citation>
          </citation>
          <citation key="ref12">
            <unstructured_citation>R. F. Liu, Verification study of probabilistic fracture mechanics analysis code for reactor pressure vessel during pressurized thermal shock, International Journal of Mechanical Engineering and Applications, vol. 10, 17-24, 2022.</unstructured_citation>
          </citation>
          <citation key="ref13">
            <unstructured_citation>Md. Sifatual Muktadit, D. Datta and A. S. Mollah, Probabilistic Fracture Mechanics Analysis of the Beltline of A PWR Nuclear Power Plant Pressure Vessel, Proc. of the 2021 International Conference on Automation, Control and Mechatronics for Industry 4.0 (ACMI), IEEE Xplore: 08 September 2021. DOI: 10.1109/ACMI53878.2021.9528105</unstructured_citation>
          </citation>
          <citation key="ref14">
            <unstructured_citation>S. M. A. S. Chapuliot, Improvement of the calculation of the stress intensity factors for underclad and through-clad defects in a reactor pressure vessel subjected to a pressurised thermal shock, International Journal of Pressure Vessels and Piping, vol. 85, 540-547, 2008.</unstructured_citation>
          </citation>
          <citation key="ref15">
            <unstructured_citation>K. Fukuya, H. Nishioka and K. Fujii, Fracture behavior of austenitic stainless steels irradiated in PWR, Journal of Nuclear Materials, 378(2), 211- 219, 2008.</unstructured_citation>
          </citation>
          <citation key="ref16">
            <unstructured_citation>D. G. Morris, Influence of carbon content on the properties of austenitic steels, Materials Science and Engineering, vol. 448, 287–294, 2007.</unstructured_citation>
          </citation>
          <citation key="ref17">
            <unstructured_citation>Adam Rylski, and Krzysztof Siczek, Issues relative to the welding of Nickel and its Alloys, Materials, 18(15), 3433, 2025. https://doi.org/10.3390/ma18153433</unstructured_citation>
          </citation>
          <citation key="ref18">
            <unstructured_citation>Ghazi Ardekani, Seyed Fazel and Hadad, Kamal, Evaluation of radiation damage in belt-line region of VVER-1000 nuclear reactor pressure vessel, Progress in Nuclear Energy, vol. 99, 96-101, 2017.</unstructured_citation>
          </citation>
          <citation key="ref19">
            <unstructured_citation>U. N. R. Commission, Evaluation of the beltline region for nuclear reactor pressure vessels, 2013.</unstructured_citation>
          </citation>
          <citation key="ref20">
            <unstructured_citation>European Nuclear Society, Reactor Pressure vessle, https://www.euronuclear.org/glossary/reactorpressure-vessel, 2023.</unstructured_citation>
          </citation>
          <citation key="ref21">
            <unstructured_citation>S.J. Zinkle and G.S. Was, Materials challenges in nuclear energy, Acta Materialia, Vol. 61, Issue 3, 735-758, 2013.</unstructured_citation>
          </citation>
          <citation key="ref22">
            <unstructured_citation>Álvaro Rodríguez, Ana María Camacho, and Miguel Ángel Sebastián, Prediction of the mechanical behaviour of cladding materials for nuclear reactor pressure–vessels based on the analysis of technological requirements, Procedia Engineering 100, 1301-1308, 2015.</unstructured_citation>
          </citation>
          <citation key="ref23">
            <unstructured_citation>I. A. E. Agency, Design and safety of reactor pressure vessels, Vienna, Austria, 2000.</unstructured_citation>
          </citation>
          <citation key="ref24">
            <unstructured_citation>U. N. R. Commission, AP600 Design Certification Document, Washington, D.C., 2000.</unstructured_citation>
          </citation>
          <citation key="ref25">
            <unstructured_citation>L. J. Douglass, Probabilistic fracture mechanics in nuclear reactor pressure vessel analysis, Journal of Pressure Vessel Technology, vol. 127, 167-173, 2005.</unstructured_citation>
          </citation>
          <citation key="ref26">
            <unstructured_citation>Donald R. Askeland, and Wendelin J. Wright, Essentials of Materials Science and Engineering, Cengage Learning, January, 2013.</unstructured_citation>
          </citation>
          <citation key="ref27">
            <unstructured_citation>Seismic Design,Bangladesh National Building Code, Dhaka, Bangladesh, 2006.</unstructured_citation>
          </citation>
          <citation key="ref28">
            <unstructured_citation>M. Hossain, Seismic hazard analysis and its impact on the nuclear reactor pressure vessel design in Bangladesh, Journal of Earthquake Engineering, vol. 24, 565-580, 2020.</unstructured_citation>
          </citation>
          <citation key="ref29">
            <unstructured_citation>IAEA. Integrity of reactor pressure vessels in nuclear power plants: assessment of irradiation embrittlement effects in reactor pressure vessel steels, IAEA NUCLEAR ENERGY SERIES No. NP-T-3.11, 2009.</unstructured_citation>
          </citation>
          <citation key="ref30">
            <unstructured_citation>Shizhong Wei, Zhu Jinhua, Liujie Xu and Long Rui, Effects of carbon on microstructures and properties of high vanadium high-speed steel, Materials &amp; Design, 27(1),58-63, 2006. DOI: 10.1016/j.matdes.2004.09.027</unstructured_citation>
          </citation>
        </citation_list>
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