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Research work of this study focuses on the seismic soil-structure interaction (SSSI) analysis of a base isolated nuclear building (NB) of the Indian pressurized heavy water reactor (IPHWR) under the action of coherent ground motion using the frequency domain approach. The seismic isolation of NB is achieved by installing lead rubber bearings (LRBs) below the raft. The base isolated NB is subjected to SSSI analysis in the frequency domain in order to take the influence of soil-structure interaction into account. LRB base isolators are numerically simulated as spring elements by assigning iterated equivalent linear stiffness. The seismic analysis of base isolated NB is performed using a conventional approach of fixed base analysis and the frequency domain approach of SSSI analysis to evaluate the effect of inclusion of SSSI on NB structures' performance. This study also evaluates the implication of modeling isolators by assigning constant isolator stiffness as well as assigning equivalent linear stiffness to account for its nonlinear behavior under seismic excitation. Seismic responses in terms of in-structure response spectra and seismic displacements at representative locations of NB substructures are studied. The outcome of the study showed (1) base isolation of NB of IPHWR is achieved successfully by bringing down spectral accelerations and relative structural displacements of NB structure at target frequencies; (2) the conventional approach of fixed base boundary condition and assigning constant isolator parameters derived at isolator design displacement underestimates the structural responses in terms of spectral acceleration, frequencies, and displacements; (3) the conventional fixed base analysis approach fails to capture the spectral amplifications along its height at the fundamental frequencies of the various structure of NB; (4) discounting the SSSI effect along with equivalent iterative isolator parameters underestimates the structural responses marginally under horizontal excitation; and (5) the SSSI effect works like an isolation in the vertical direction and leads to reduced spectral accelerations in the vertical direction. [ABSTRACT FROM AUTHOR]

The Liquid Phase Catalytic Exchange (LPCE) process plays a pivotal role in the separation of hydrogen isotopes, particularly in applications such as tritium removal in heavy water reactors. Effective separation is crucial for maintaining reactor safety and efficiency. In this study, the optimal operating conditions for the LPCE process were determined through orthogonal experiments and validated in different hydrogen isotope systems. The experiments investigated key operational parameters, including the filling ratio of catalyst to packing (FR), operating temperature (T), superficial gas velocity (V), and gas-to-liquid flow rate ratio (λ), using a robust L16 orthogonal experiment design. The results indicated that V and FR had the most significant effects on the height equivalent to a theoretical plate (HETP), while λ exhibited the greatest impact on dedeuterization efficiency (DE). The optimal conditions obtained were V = 0.1 m/s, FR = 1:2, T = 70 °C, and λ = 2.5. Furthermore, the reproducibility of the optimal conditions was verified in LPCE columns with varying diameters (1.5 cm, 2.5 cm, 4.5 cm). Additionally, the findings were applied to both H-D and D-T separation systems, demonstrating consistency in mass transfer efficiency and validating the applicability of the optimal conditions in different hydrogen isotope separations. This research provides critical insights for optimizing tritium removal systems in heavy water reactors, contributing to enhanced reactor safety and performance. [ABSTRACT FROM AUTHOR]

Flux flattening and power uprating of large heavy water power reactors (HWRs) are demonstrated as an application of an accelerator-driven photoneutron source (ADS) in the ADS-CANDU concept where an array of electron linear accelerators is configured around the periphery of a subcritical CANDU-6 core. The localized ADS generated through (e−,γ,n) reactions in the HWR lattice perturbs the reactor power distribution by increasing the power of low-power bundles and depressing the power at the core center relative to the fundamental mode power distribution. Gross power uprating is feasible when the system is near critical, but the ADS array consumes tens of megawatts electric exceeding the power gained by a factor of more than 2 for the conservative ADS performance specifications assumed in the analysis. Several important challenges of fixed-source Monte Carlo simulations of near-critical multiplying media are investigated including severe load imbalance issues with distributed-memory parallel computing architecture and correlated local tallies in nonanalog (implicit absorption) Monte Carlo radiation transport. All subcritical fixed-source simulations in the study readily exceed the default random number stride used in most production Monte Carlo codes, and the stride exceedance causes both bias in local tally results (bundle powers) and spatial autocorrelation of these errors/biases in the large core. A legacy stride exceedance is critically reviewed, and the conclusions and subsequent interpretations of those conclusions are rejected. Several classes of radiation transport Monte Carlo problems are likely to be susceptible to stride exceedance, and this issue needs to be promptly addressed by the Monte Carlo analyst and code developer communities. [ABSTRACT FROM AUTHOR]

The routine practice accompanying the operation of nuclear facilities involves the discharge of radioactive effluents from nuclear power plants (NPPs). Regulation of this discharge to the environment hinges on three criteria: radioactivity concentration, public dose, and radioactivity. Among these, radioactive carbon‐14 holds particular significance as it possesses an extensive half‐life of 5730 years, making it a primary source of radiation dose to communities residing around NPPs. In Korea, the monitoring of carbon‐14 discharges from pressurized water reactors (PWRs) in gaseous effluents has been ongoing since 2012, whereas before 2012, monitoring exclusively focused on carbon‐14 discharges from pressurized heavy water reactors (PHWRs). Analysis of carbon‐14 discharges from Korean PHWRs indicates that their emission constituted less than 1% of total radioactive effluents over the past two decades. In the context of Korean PWRs, carbon‐14 discharge monitoring was absent from 2002 to 2011, resulting in an absence of data regarding such discharges during that period. After introducing carbon‐14 discharge monitoring in gaseous effluents from Korean PWRs, emissions from 2012 to 2021 contributed 3% of the total gaseous effluents. These findings indicate that despite being the primary contributor to public dose, carbon‐14 discharges from NPPs constitute a minor portion of the radioactive effluent discharge. [ABSTRACT FROM AUTHOR]

The heavy water reactor nuclear fuel channel is loaded with nuclear fuel bundles, therefore, its structural integrity is important. Since the fuel channel structure is affected by radiation irradiation as well as flow-induced vibration due to heavy water coolant flow, it is necessary to monitor and diagnose its structural health condition in real time. Therefore, measurement systems of neutron noise and structural vibration are utilized in heavy water reactor system. In this paper, we introduce a research regarding an AI (artificial intelligence) system that can monitor the presence of mechanical defects at supports of the nuclear fuel channel, and accurately locate the failure area using vibration data of the nuclear fuel channel and CNN (convolution neural network). Vibration data of the fuel channel for normal and abnormal conditions were generated by fluid-structure interaction analysis. CNN and AI systems were trained to recognize different vibration characteristics of normal and various failure modes using images of short time Fourier transform. It was confirmed that the proposed AI accurately identifies the presence of mechanical defect of support structures and its location, even when the mass flow rate of heavy water varies. The results of this research can also be applied to develop AI system for fault diagnosis of nuclear fuel channels based on neutron noise data obtained by neutron flux detectors installed in the actual heavy water reactor so that the structural integrity of the heavy-water nuclear power plant system could be ensured more reliably. [ABSTRACT FROM AUTHOR]

Nuclear reactors will face the problem of decommissioning at the end of their operating life due to the high radioactivity of reactor components and environmental safety considerations. The Heavy Water Research Reactor (HWRR) is the first large-scale research reactor to be decommissioned in China. The second phase of HWRR decommissioning involves the main components in the reactor block, so the radiation source terms and the radioactive waste level need to be evaluated before demolition and disposal. Based on the operating history, three-dimensional geometry, materials, and other information of the HWRR, the activity of radionuclides in the main components of HWRR is calculated and analyzed, and the MCNP/ORIGEN coupling scheme is utilized for theoretical analysis. The theoretical results indicate that 14C, 54Mn, 55Fe, 60Co, 63Ni, and 152Eu are the main radioactive nuclides. The total activity of radioactive nuclides was 2.36E + 15 Bq at the end of 2007, 4.27E + 13 Bq at the end of 2021, and 1.83E + 13 Bq at the end of 2025. Furthermore, local sampling and radiometric analyses based on the HPGe gamma-ray spectrometer are also performed to verify the theoretical results, the ratio of theoretical activity values to the measured activity of the experimental sample is within 2.5 times, so the theoretical results are conservative. According to the classification standards for radioactive waste, the inner shell, outer shell, cooling water tank, sand layer, and heavy concrete shielding layer are all low-level waste. These results and conclusions can serve as a reference for the second phase decommissioning of the HWRR and the subsequent disposal of radioactive waste. [ABSTRACT FROM AUTHOR]

The article focuses on India's ambitious plans to expand its nuclear energy capacity in response to growing energy demands and climate goals. Topics include the role of nuclear power as a reliable and clean baseload energy source, the development stages of India's nuclear program including the use of pressurized heavy water reactors and fast breeder reactors, and the challenges in scaling up to meet the 100 GW target by 2047.

Currently, over 100 nuclear power units globally have been in operation for more than 40 years. Hindered by the limitations of computer technology at the time, these nuclear facilities lack detailed electronic drawings. Activities such as equipment replacement and process circuit system modifications during operation result in discrepancies between paper drawings and actual conditions. Given the complexity and irreversibility of nuclear facility decommissioning activities, virtual simulation technology is often employed before the decommissioning process begins to assist in designing and validating decommissioning plans. Consequently, the creation of high-precision 3D models is crucial for subsequent decommissioning designs. Through innovatively utilizing laser-scanning 3D model reconstruction technology in the reconstruction of the model of China's first heavy water research reactor undergoing decommissioning, this paper provides an overview of the process of laser-scanning 3D model reconstruction and its application in reconstructing the heavy water research reactor model. Using a 3D laser scanner, four decommissioning areas of the heavy water research reactor, including the reactor building, secondary water pump room, ventilation center, and low-level radioactive wastewater storage tank area, were subjected to 3D laser scanning. The acquired point cloud data from 572 scanning stations were processed using point cloud processing software for denoising, stitching, and triangulation. The triangulated model was then imported into modeling software for 3D reconstruction, ultimately establishing a digitalized model of the heavy water research reactor suitable for subsequent decommissioning simulation and design. [ABSTRACT FROM AUTHOR]

Zirconium alloy (Zircaloy-4) is widely used for fuel cladding in the Canadian Deuterium Uranium (CANDU) heavy water reactors. However, degradation due to waterside corrosion can limit the in-reactor design life of the nuclear fuel. For this reason, efforts must be made to understand the mechanisms of corrosion and to mitigate its effects. The purpose of the experimental research related to this work consists of the assessment of corrosion of the CANDU fuel cladding material. The paper presents the results obtained by in-situ monitoring of the Pressurized Heavy Water Reactor (PHWR) water chemistry effect on fuel cladding corrosion, Zircaloy-4. Optical metallographic and Scanning Electron Microscopies (SEM), as well as X-ray diffraction (XRD) analysis, were used to evaluate the corrosion behavior of the fuel cladding material, Zircaloy-4 coupons, exposed in a Primary Heat Transport System (PHTS) autoclave system. The obtained results are useful for characterizing the corrosion behavior of Zircaloy-4 coupons exposed in the autoclave system at Cernavoda NPP for long periods. [ABSTRACT FROM AUTHOR]

Kori Unit 1, which was permanently shut down in 2017, is expected to be the first decommissioned commercial nuclear power plant (NPP) in Korea. Operating NPPs designate radiation‐controlled areas (RCAs) to protect workers from radiological hazards and provide appropriate measures. Decommissioning RCAs (DRCAs) in NPPs being decommissioned should also be designated by considering the radiological characteristics of workplaces and exposure routes of decommissioning workers, such as external and internal exposure. However, the criteria for the DRCAs in decommissioning Kori Unit 1 are presently the same as those during normal operation, according to the public‐hearing material of the final decommissioning plan for Kori Unit 1. This study analyzed criteria for RCAs in all Korean NPPs to propose criteria for DRCAs in decommissioning NPPs. Analysis results show that RCAs are classified into four, five, six, and eight zones in Korean pressurized water reactors with respective external radiation dose rates. The RCAs in Korean pressurized heavy water reactors are classified into six zones with external dose rates and derived air concentrations of tritium. This study proposes three classifications of DRCAs to keep occupational exposure as low as reasonably achievable while considering potential exposure routes, that is, external exposure, airborne contamination, and surface contamination areas. First, the external exposure areas are classified into several zones according to the levels of external radiation dose rates. Second, the airborne contamination areas are designated to prevent internal exposure to workers from the inhalation of radioactive material in the air during cutting and demolition. Last, the surface contamination areas are designated to minimize the skin contamination of workers as radioactive dust in the air is deposited on the surface of facilities and equipment over time. [ABSTRACT FROM AUTHOR]

Hard bremsstrahlung produced during the deceleration of fission-product beta rays in nuclear fuel is proposed as a source of delayed photoneutrons (PNs) in heavy water reactors. Electron Gamma Shower (EGS5) code simulations of coupled electron-photon transport in seven fuel element geometries immersed in an infinite heavy water medium confirm high-energy beta rays produce sufficient bremsstrahlung yields with photon energies greater than the D(γ,n)1H reaction threshold such that the beta-ray contribution to an isotopic PN yield can be comparable to or greater than yields from hard gamma rays emitted during the isomeric transition of the daughter, especially for some short-lived fission products with high-intensity direct-to-ground-state beta transitions where the beta ray and antineutrino carry away the majority of the Q-value. Some fission products that do not have hard gamma rays in their decay schemes are in fact PN precursors due to the beta-ray-bremsstrahlung contribution. The lack of evaluated nuclear data for many short-lived fission products, data reliability issues of fission products with decay scheme data, cross-section library effects, and possible decay-chain/parent-feeding phenomenon are affecting the accuracy of PN group parameters derived from a small number of legacy experiments. The reanalysis of two legacy PN experiments supports the existence of a very-short-lived direct-delayed neutron group. [ABSTRACT FROM AUTHOR]

During the conventional operation, nuclear power plants discharge waste water containing amount of radioactivity, one of main ingredient is the artificial radionuclide tritium. In this paper, the chemical characteristics and the level of tritium in the environment are studied. The monitoring medium and results of tritium in eight typical nuclear power plants are described. According to the monitoring data, the emission of tritium in heavy water reactor is relatively high, while that in pressurized water reactor is basically at the background level. Tips on topics related to strengthen the monitoring of radiation environment around the nuclear power plants have noted. [ABSTRACT FROM AUTHOR]

All light or heavy water reactors fabricated with carbon steels suffer from flow-accelerated corrosion (FAC). The FAC degradation of SA106B with different flow velocities was investigated in terms of microstructure. As flow velocity increased, the major corrosion type changed from general corrosion to localized corrosion. Severe localized corrosion occurred in the pearlite zone, which can be the prior location for generating pits. After normalizing, the improvement in microstructure homogeneity reduced the oxidation kinetics and lowered cracking sensitivity, causing a decrease in FAC rates of 33.28%, 22.47%, 22.15%, and 17.53% at flow velocity of 0 m/s, 1.63 m/s, 2.99 m/s, and 4.34 m/s, respectively. Additionally, localized corrosion tendency was decreased by reducing the micro-galvanic effect and tensile stresses in oxide film. The maximum localized corrosion rate decreased by 21.7%, 13.5%, 13.8%, and 25.4% at flow velocity of 0 m/s, 1.63 m/s, 2.99 m/s, and 4.34 m/s, respectively. [ABSTRACT FROM AUTHOR]

The article offers information on the 89th Annual Meeting of the Indian Academy of Sciences (IASc), which was held in Bengaluru.

In this technical note, an analysis of an integral experiment of the Advanced Heavy Water Reactor (AHWR) Critical Facility (CF) with a diffusion-based Monte Carlo (MC) method is discussed. In this method, the diffusion kernel is converted into probabilities per unit time for tracking the particle in the problem domain. The diffusion-based MC method is coupled with a time-dependent MC algorithm developed earlier and has been used for space-time simulations in neutron multiplication assemblies. Kinetics simulations are best solved using a transport MC route, but this requires long computational time. The diffusion-based MC method provides a faster solution in such space-time simulations. Most of the space-time kinetics studies and benchmarks are based on diffusion theory, and there are very few transport theory or MC benchmarks. Thus, the diffusion-based MC facilitates exact comparison with the large number of diffusion theory benchmarks. The efficacy of this method was tested earlier by comparison with the results of realistic space-time kinetics benchmarks based on diffusion theory methods involving multiregion reactors and detailed energy dependence. Comparison of our results with these benchmarks has shown satisfactory agreement. As a step toward more detailed benchmarking, the ability and accuracy of this method are tested on the recent experiment done in the AHWR CF. The integral experiments with one thoria-based mixed oxide experimental fuel assembly in the core of the AHWR CF were analyzed with this method and were compared with the observed experimental values. The experiments consisted of measurement of the critical height and worth of shut-off rods (SORs) with the experimental fuel assembly placed at different lattice locations. Neutron count rates as a function of time after reactor trip for estimation of the worth of the SORs were also simulated, and the results are found to be in good agreement with the observed values. [ABSTRACT FROM AUTHOR]

Plastic anisotropy is observed during plastic flow in samples from the axial and transverse directions of Zr-2.5Nb pressure tubes used in CANDU 1 1 CAN ada D euterium U ranium - Pressurized Heavy Water power reactors. nuclear reactors tested in uniaxial tension. Plastic anisotropy was also measured in shear by testing in torsion 'mini' pressure tubes from the same material. Room temperature results from these experiments were analysed using a visco-plastic self-consistent model that takes into account the crystallographic texture of the material and allows for the single crystal work hardening behaviour to be described by means of a deformation law specific to each slip system. The visco-plastic self-consistent model was used to derive: (i) the evolution with strain of the critical resolved shear stress values consistent with prismatic, basal and pyramidal dislocation glide, (ii) the evolution of slip system activity as a function of strain, and (iii) the values of Hill's plastic anisotropy coefficients that are consistent with the observed anisotropy of yielding and their dependence on accumulated strain. The model also allows for the prediction of the components of the flow stress tensor that cannot be measured experimentally and their dependence on the work hardening behaviour of pressure tube material. Moreover, the yield surface after different amounts of plastic strain was calculated using the visco-plastic self-consistent model, compared to the one that results from using the Hill's anisotropy coefficients. Our work exposes the limitations of the Hill ellipsoid for describing plastic yield when microstructural evolution is present. [Display omitted] [ABSTRACT FROM AUTHOR]

An extensive experimental investigation was performed to study the oxygen embrittlement of the Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under simulated Loss-of-Coolant Accident (LOCA) conditions. Zircaloy fuel cladding experiences creep and corrosion simultaneously during service and LOCA conditions. Zircaloy-4 fuel pins were pre-oxidized to attain different oxide layer thicknesses, achieving in-service conditions. These pre-oxidized tubes were then subjected to burst tests in the steam environment to mimic the LOCA scenario. The present study aims to improve the understanding of the effect of oxidation on the cladding microstructure and the mechanical response of the fuel pin in a LOCA scenario by accounting for the cross-influence, during transient heating, of oxidation and deformation on the behavior of the clad in the LOCA domain. The oxide layer morphology in pre- and post-burst samples was studied using FESEM, XRD, and Raman spectroscopy. In some cases, the inner oxide layer grew faster than the outer oxide layer when the fuel pin was heated in steam during the burst test. The evolution during transient heating of radial and circumferential crack growth in the oxide layer and the occurrence of delamination facilitated faster oxygen and hydrogen uptake. The hydrogen uptake in pre and post-burst samples was related to the oxygen uptake. The hydrogen concentration increases with the oxygen concentration in the pre-oxidized samples. Small oxygen and hydrogen concentrations were found in the post-burst as-received samples due to the formation of a protective oxide layer. The high-temperature oxide layer was formed at extremely high heating rates. [Display omitted] • Studied cross-influence of oxidation and deformation induced during simulated LOCA. • Performed burst tests on as-fabricated and pre-oxidized zircaloy-4 claddings. • Different heating rates and internal pressures are selected to mimic LOCA. • Post-burst characterization is performed using FESEM, XRD, and Raman spectroscopy. • Hardness and elastic modulus are evaluated using nanoindentation. [ABSTRACT FROM AUTHOR]

The Taiwan Research Reactor (TRR) is a research facility in Taiwan. TRR is a heavy water reactor designed based on the National Research Experiment (NRX) with output thermal power 40 MW. On January 3, 1973, it reach criticality. TRR had been operated for 15 years and was permanently shut down in early 1988. TRR uses heavy water moderation and light water cooling. Hence, some tritium was created when deuterium captures a neutron. The inside of the Calandria has been washed with clean water after shutdown. After draining the water in the interior, it is inevitable that some residual water would remain at the bottom of the Calandria. Air sampling and analysis techniques were employed to detect and quantify tritiated water vapor. The investigation of tritiated water vapor in the TRR Calandria proved potential tritium contamination risks. Adopting the vapor condensation drying technique proved effective in removing tritiated water vapor from the TRR Calandria. After the removal of tritiated water vapor, the air in the Calandria no longer contains rich concentration of tritiated water vapor. That will help to ensure a safe environment for decommissioning operations. [ABSTRACT FROM AUTHOR]

To ensure a reliable shutdown of the proposed advanced heavy water reactor (AHWR) of Indian origin during emergency conditions, it is imperative to use an increased concentration of liquid neutron poison for injection into the moderator system due to its higher reactivity. The sudden injection of this liquid poison may cause a surge in radiolytic product (H 2 O 2 and H 2) yield, posing safety risks. If H 2 /D 2 yield exceeds 4 % (v/v), a combustible mixture with oxygen can be formed and presence of oxidizing H 2 O 2 can enhance corrosion. Hence, accurate estimation of H 2 O 2 and H 2 is imperative from the reactor safety point of view. We investigated gadolinium and boron (two liquid neutron poisons) at varying concentrations through high energy electron beam and gamma radiolysis. The radiolytic yields from both the liquid neutron poisons of equivalent concentrations at a constant dose and operating moderator conditions in electron beam and gamma radiolysis were compared. The molecular product yield was observed to be concentration and dose-rate-dependent for both the neutron poisons. Electron radiolysis exhibited a higher yield of H 2 O 2 , whereas gamma radiolysis resulted in a greater H 2 yield. In electron radiolysis, the H 2 O 2 yield decreased with increasing B concentration. These results give an overall estimation of radiolytic products for a particular dose rate, total dose and effect of head space volume over liquid irradiation zone from these two liquid poisons of a particular concentration. These insights will aid in selecting suitable poisons and managing their effects effectively to avoid hydrogen deflagrations after emergency shutdown. [ABSTRACT FROM AUTHOR]

Passive heat removal systems (PHRS) are indispensable additional safety features in nuclear reactors to deal with station blackout (SBO) conditions. These systems are based on coupled natural circulation loops (C-NCL) connected in series configuration. Further, these system's design, operating conditions, working fluid, heat removal capacity, etc., are specific to the reactor type. In the present work, the operational performance of such first of a kind PHRS of 700 MWe Indian pressurized heavy water reactor (IPHWR) is investigated. For this purpose, a coupled two-phase NCL code is developed, validated and employed. This PHRS consists of two NCLs connected in HHVC-VHVC (horizontal heater and vertical cooler - vertical heater and vertical cooler) configuration. In this study, initially, the full power reactor thermal-hydraulic conditions are established, subsequently, the performance of the PHRS under SBO is investigated. The analyses demonstrated that the system is adequate to effectively eradicate the decay heat by establishing single and two-phase natural circulation (NC) flows of ∼4.5 % and ∼5.6 % of rated conditions in primary and secondary transport loops, respectively. The maximum PHRS capability is predicted to be ∼72 MWth which is more than 3 % of full power. The transients of flow, pressure, energy transfer and heat sink thermal conditions are analyzed when the system changes from forced to NC. Further, the influence of system conditions such as loop diameter, loop height, sink volume, sink temperature and heat exchanger area on PHRS performance was predicted and the changes in system parameters were analyzed. The vital parameters influencing the PHRS performance are identified based on sensitivity analyses. This investigation provides significant insight into the flow and heat transfer characteristics of two-phase and coupled NCLs in the PHRS. • Developed a C-NCL model and applied it to a passive heat removal system (PHRS). • Investigated the performance of first of a kind PHRS of an IPHWR under SBO. • Thermal-hydraulics of PHRS are predicted from forced to natural circulation. • Adequacy of PHRS for IPHWR to remove decay heat under SBO is demonstrated. • Effect of system conditions on PHRS is studied and identified key parameters. [ABSTRACT FROM AUTHOR]

[Display omitted] • Burst testing of IPHWR zircaloy-4 claddings at 55–115 K/s and 15–45 bar. • Developed burst stress correlation as a function of oxygen pickup and burst temperature. • Developed burst criterion model and validated it with experimental data. • Studied the effect of various parameters on the burst behavior of zircaloy-4 cladding. This paper presents a detailed investigation of the creep and rupture behavior of Zircaloy-4 (Zry-4) fuel claddings used in Indian Pressurized Heavy Water Reactors (IPHWR) under simulated Loss-Of-Coolant Accident (LOCA) conditions. Fuel claddings are pre-oxidized in the steam environment at 500 °C to mimic the oxygen pickup during normal reactor operation. The burst tests are then performed on these preoxidized tubes at different heating rates (55–115 K/s) and internal overpressures (15–45 bar) in the steam environment, creating LOCA-like scenario in a high burnup condition, wherein the claddings further oxidize while undergoing deformation and rupture. The burst stress correlation is developed for IPHWR claddings from the obtained burst temperature and oxygen concentration data. A burst criterion model is developed by solving available creep rate, oxidation rate, and phase transformation equations simultaneously to study the effect of various parameters on burst characteristics of the fuel cladding. The proposed burst criterion model agrees well with the present and previous experimental burst data. Also, the ballooning progression predicted from the model is validated with the present and previous experimental data. In addition, the effect of available correlations for the creep rate, phase boundary temperature, and oxidation factor on the burst characteristics has been presented. [ABSTRACT FROM AUTHOR]

LT21 a type of aluminum alloy used for the irradiation channel of the first heavy water research reactor (HWRR) in China. Studying the mechanical property of irradiated LT21 aluminum under actual service conditions is essential for evaluating its application property. In this paper, tensile specimens of irradiated LT21 were manufactured from the decommissioned irradiation channel of an HWRR; then, tensile tests were carried out, and then the fracture surfaces were observed. The effect of neutron irradiation on tensile behavior and the failure mechanism was analyzed by comparing the result of irradiated and unirradiated LT21 specimens. The results show that, with the thermal neutron flux increasing to 2.38 × 1022 n/cm2, the YS gradually increased from the initial 158 MPa to 251 MPa, the UTS increased from 262 MPa to 321 MPa, and the elongation decreased from 28.8% to about 14.3%; the brittle fracture of the LT21 specimen appeared after irradiation, and the proportion of brittle fracture increased as the neutron fluence increased; the nanophase structures, with a size of less than 50 nm, were precipitated in the LT21 aluminum alloy after neutron irradiation. Transmutation Si is presumed to be the main cause of the radiation effect mechanism of LT21. [ABSTRACT FROM AUTHOR]

In the current work, the high-temperature deformation behaviour of Zr-2.5Nb alloys used in pressurized heavy water reactors has been studied. Uniaxial tensile tests were carried out for both longitudinal and transverse specimen orientations in the temperature range of 25–800 °C at a nominal strain rate of 1 × 10–4 s−1. The transverse samples exhibited higher strength and lower ductility than the longitudinal ones at all temperatures. Percentage elongation of 406% and 386% was observed for the longitudinal and transverse samples, respectively, at 800 °C. The variation of tensile properties has been correlated over a wide range of temperatures by empirical equations, which will be beneficial for the computation of the elevated temperature deformation behaviour during the loss of coolant accident scenario in the reactors. TEM studies revealed the presence of Nb-enriched precipitates at the grain boundaries at elevated temperatures. The high value of strain rate sensitivity (m) obtained from the strain rate jump test in the temperature range of 600–800 °C confirmed the superplastic behaviour of this alloy in the region. The m-value and activation energy indicated that grain boundary sliding was the primary mechanism during superplastic deformation, accommodated by a combination of lattice and grain boundary diffusion. [ABSTRACT FROM AUTHOR]

The Heavy Water Board (HWB), an industrial unit of the Department of Atomic Energy (DAE), Government of India is engaged in the production of heavy water and specialty materials required by DAE for supporting the Indian Nuclear Power Programme (INPP). HWB through its four operating plants has ensured self-sufficiency in the availability of heavy water for pressurized heavy water reactors and has also exported heavy water. The mandate of HWB has been enlarged to include the production of specialty materials for DAE. This includes the development of technology for the production of solvents required in pursuit of a closed nuclear fuel cycle, development and deployment of technology on an industrial scale for the production of special materials such as nuclear-grade sodium, enriched boron compounds and boron carbide pellets for fast breeder reactors. Development and deployment of new processes, technologies, systems and equipment in areas of relevance to DAE, viz. recovery of value metals like cobalt, gallium, rare earth materials, etc. from the secondary resources is also a part of the enlarged mandate of HWB. The Board is also deploying spin-off technologies and providing consultancy services in the field of energy conservation, engineering and project management. [ABSTRACT FROM AUTHOR]

Considering the energy resource profile in the country, Homi Bhabha underscored the need for the development of atomic energy, advocated the pursuit of a closed fuel cycle and selected the reactor system to be adopted by India for large-scale deployment. Following the vision outlined by him, India selected pressurized heavy water reactors (PHWRs) of smaller capacity (i.e. 220 MW plants) for large-scale commercial deployment of nuclear energy. Over time, the country developed expertise in this technology with the construction of several 220 MW plants and then raised the capacity to 540 MW. This experience, expertise and indigenization have led to the development of a state-of-the-art 700 MW PHWR, and India is now set to install several such reactors along with light water reactors in technical collaboration with foreign vendors, to increase the nuclear-installed base in the country. Technologies to reprocess spent nuclear fuel and manage high-level waste have also been developed and deployed. This article presents a brief outline of the present status of nuclear power development in India. [ABSTRACT FROM AUTHOR]

• Subcooled flow boiling under high mass flux and heat flux is studied. • Local heat transfer coefficient (HTC) & pressure drop characteristics are studied. • IR thermography is used for outer wall temperature measurement. • Effects of operating & geometric parameters on HTC & pressure drop are studied. • Non-dimensional correlations for local HTC and pressure drop are proposed. Subcooled flow boiling of water is widely observed in high heat flux and high mass flux (HHHM) cooling applications such as heat exchangers, refrigeration equipment, boiler tubes and nuclear reactor core fuel channels in pressurized heavy water reactors (PHWR). In this study, the focus is on investigating the local heat transfer coefficient (HTC) and pressure drop in a horizontal tube experiencing subcooled boiling of water under low pressure and HHHM conditions. The study encompasses different geometrical parameters such as tube diameter (5.5 mm , 7.5 mm , 9.5 mm and 12 mm) and length (550 mm for each of the tubes). The operating parameters that are varied include mass flux (248–2000 kg/m 2 . s) and heat flux (0–1837 kW/m 2 ). Infrared thermography is used to measure the local wall temperature. A non-dimensional correlation for the diabatic pressure drop ratio (ratio of diabatic pressure drop to adiabatic pressure drop) as a function of Jakob number (Ja), Boiling number (Bo) and diameter ratio is developed. Subcooled boiling pressure drop ratio for 5.5 mm , 7.5 mm and 9.4 mm diameter tubes is 2.23 which is independent of diameter. A correlation for the two phase local HTC during subcooled flow boiling conditions as a function of Ja , Bo and Prandtl number (Pr) is also developed. [ABSTRACT FROM AUTHOR]

• We developed a Decommissioning Waste Assessment Module (DWAM) for HWRs. • DWAM is an advanced software tool designed to optimize the decommissioning process. • DWAM can serve as a simulation tool to select the optimal decommissioning process. In this study, we developed a Decommissioning Waste Assessment Module (DWAM) for Heavy Water Reactors (HWRs). To evaluate the waste amount of radioactive structures that make up a significant portion of decommissioning waste, we establish a comprehensive DB containing detailed information on the volume, weight, material type, and radioactivity of each decommissioning waste segment. Decommissioning waste assessment algorithms were developed and applied in the proposed module to optimize the calculated results, and their performance was evaluated to ensure accuracy and reliability. The DWAM currently being developed for the Wolsong Unit 1 commercial pressurized heavy water reactor is a technology that has the potential to be utilized not only for domestic reactors but also for reactors worldwide. In addition, the technology is expected to contribute to the development of technology related to the decommissioning of both radioactive and non-radioactive structures. [ABSTRACT FROM AUTHOR]

• Interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is reported. • Dominance of convection heat transfer is observed for initial exposures of the fuel channels. • Advanced stages of fuel channel exposures are observed to be dominated by radiation heat transfer. • Convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation from Zr-Steam interaction. • Heat-up of fuel channels is restricted to central regions of the exposed core due to degraded steam cooling and higher-powered fuel channels. • Fuel channel disassembly is observed beyond eight rows of fuel channel exposure. • Fuel channel length equivalent to nearly 7–8 fuel bundle lengths undergo disassembly. • Calandria vault water is observed to undergo film boiling. A three-dimensional steady-state thermal–hydraulic analyses of a large sized Pressurized Heavy Water Reactors (PHWRs) with varying levels of exposure is carried out for degraded core heat transfer assessment and presented in this paper. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in PHWRs. Such exposure of fuel channels leads to their heat-up, governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer and thermo-chemical interactions among the core constituents. In order to understand such complex behavior and plan the severe accident management guidelines, numerical analyses of the exposed core are of necessary. This paper illustrates the influence of the temperature and velocity profiles vis-à-vis different regions of the exposed core on the fuel channel heat-up. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is reported. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, the advanced stages of fuel channel exposures are observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation from Zircaloy-made calandria tube surface and steam interaction. This leads to supplemental power addition for fuel channel heat-up. Moreover, it is interesting to note that the heat-up of fuel channels is restricted to central regions of the exposed core due to degraded steam cooling and higher-powered fuel channels. It is worth noting that fuel channel disassembly is observed in the advanced levels of fuel channel exposures. Moreover, fuel channel length equivalent to nearly 7–8 fuel bundle lengths undergo disassembly depending upon the fuel channel elevation and location in the core. The calandria vault water is observed to undergo film boiling at the onset of fuel channel disassembly. [ABSTRACT FROM AUTHOR]

A simulation model is presented for the burnup effect of rhodium self-powered neutron detectors (Rh-SPNDs) using the Monte Carlo (MC) method. Calculated neutron sensitivities at different burnup steps are compared with the experimental measurements from the complete burnup irradiation experiments performed in a heavy water reactor (HWR). Both the simulated and experimental burnup curves of Rh-SPNDs are plotted and shows very small deviations, which proves the accuracy and reliability of the simulation methods. [ABSTRACT FROM AUTHOR]

[Display omitted] • The corrosion was observed as a weight loss due to Fe dissolution activity. • The corrosion rate was found to be 1.3 times higher with the increasing OA. • The intergranular corrosion was observed to the CS surface. The corrosion behavior of Carbon Steel (CS) SA106 Gr. B in an organic condition of oxalic acid (OA) was elucidated under the anticipated conditions of a Pressurize Heavy Water Reactor (PHWR) during the chemical decontamination. CS specimens were exposed in the acidic medium for 18 h at 95 °C to evaluate the corrosion activity. Atomic Absorption Spectrometry (AAS) was employed to analyze the dissolved Fe concentration and found to be 276.5 and 401.3 ppm in 10 mM and 20 mM OA solution, respectively. Corrosion rates were assessed using the weight loss method. The CS specimens dissolved in 20 mM exhibited a corrosion rate of 6.3 mm/y, while that dissolved in 10 mM OA has 4.6 mm/y. The corrosion products were characterized by Electron Probe Microanalysis (EPMA). The intergranular corrosion (IGC) was observed prior to the formation of corrosion products and oxalate passivation, potentially providing pathways for contaminant diffusion into the bulk substrate. [ABSTRACT FROM AUTHOR]

• CFD modelling undertaken for pressurized heavy water reactor fuel channel conditions. • Developed new wall boiling model for predicting CHF based on liquid film thickness. • Assessed model predictions against CHF tests for single-element configuration. This paper uses Eulerian multiphase computational fluid dynamics (MCFD) simulations to predict the critical heat flux in a simplified CANDU test geometry for typical pressurized heavy water reactor (PHWR) conditions. A diffused liquid film-based wall dryout model is proposed based on the physical description of the dryout phenomenon. Given the adopted ensemble-averaged two-phase framework, this proposed model can represent the film dryout physics but requires some calibration. Simulations are performed for three calibration cases at 9 MPa outlet pressures to determine the appropriate diffused liquid film layer thickness to accurately predict the experimental data for the test geometry. This calibrated layer thickness is then applied to a number of assessment cases for various axial heating profiles, mass fluxes, inlet subcooling levels, and outlet pressures. The accuracies of the assessment simulations are generally within ±11 % of the experimental measurements, with a few exceptions. Specifically, the errors for the cases at lower outlet pressures (6 MPa) and mass fluxes (≲ 2000 kg/m2-s) ranged from about −19 % to –33 %. The lower pressure cases indicate that a separate calibration may be required at low pressures, while the larger errors in the lower mass flux cases suggest that some improvement of the boiling closures is needed to account for conditions at high average outlet quality. [ABSTRACT FROM AUTHOR]

• Break opening time assumption in LBLOCA Large Break Loss of Coolant Accidents Analysis. • Demonstrating the efficiency of the second shutdown fast boron injection system during the accident analysis. • Dynamic Break Opening development for the analysis. In the frame of the licensing process of the Atucha II PHWR (pressurized heavy water reactor) the BEPU (Best-Estimate Plus Uncertainty) approach has been selected for issuing Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. In case of a LOCA (Loss of Coolant Accident) the assumption of instantaneous opening time is too conservative, especially for this kind of reactor with a positive coolant void reactivity coefficient. Analysis was performed to assess the ability of Atucha II nuclear power plant's fast boron-injection system to prevent overheating of the fuel elements in the event of a Beyond Design Basis Accident (BDBA). The goal of the Analysis was to develop a Dynamic Break Opening which was used as the input for the LOCA analyses. The results show that the fast boron injection can effectively shut down the reactor, thereby preventing the overheating of the fuel elements. [ABSTRACT FROM AUTHOR]

Safety-critical systems (SCSs) of nuclear power plants (NPPs) are being designed and developed to meet high dependability requirements. Fault tree analysis (FTA) is widely used for risk and reliability analysis of NPPs. However, fault trees (FTs) are static and have only limited capability to represent dynamic systems. FTA is also not capable of modeling non-binary logic and or modelling the system's evolution in time. Dynamic reliability methods are being developed to deal with such limitations. Time series Markov chains and dynamic flowgraph methodology are the dynamic reliability methods alternate to traditional FTA, which can be used for the system performance analysis. In this article, Time series Markov chains and DFM methods, for the system reliability predictions for the SCS of NPP are compared. The benefits of the proposed method are brought out to the traditional methods. The approach is applied on passive residual heat removal system of pressurized heavy water reactor under the station blackout scenario. [ABSTRACT FROM AUTHOR]

Large nuclear reactors operating in the thermal spectrum are prone to both global and regional oscillations in power due to variation of 135Xe concentration. These power oscillations are self-stabilizing up to a certain operating power level, beyond which spatial power control becomes necessary for suppressing these oscillations. Especially for large pressurized heavy water reactors (PHWRs), which are natural uranium–fueled reactors using heavy water as coolant and moderator, the modes of xenon instabilities decide the extent and scheme for spatial power control. In this paper, the effect of spatial control on the bifurcation characteristics is demonstrated using a two-region model. The error signal for movement of the reactivity device has a global component for bulk power control and a local component for regional power control. The amount of regional power control determines the power level at which the spatial xenon oscillations stabilize. Using bifurcation analysis, it is found that in case of limited regional control, both supercritical and subcritical Hopf bifurcations exist, whereas in the case of increased regional control only supercritical Hopf bifurcations exist. However, these supercritical Hopf oscillations are due to time lag in control and have short timescales and lower amplitudes as compared to xenon oscillations. Hence, a proper choice of spatial control enables a PHWR to operate at rated full power capacity without any spatial Xenon instability. [ABSTRACT FROM AUTHOR]

Based on this observation, 4500 epoch is selected as the optimum epoch number for the present ANN analysis for predicting Pu IC. 4.2 Optimization of ANN configuration There is no thumb rule for designing the ANN structure based on the input - output data dimension. Keywords: artificial neural network; gamma spectroscopy; high burn-up; plutonium EN artificial neural network gamma spectroscopy high burn-up plutonium 323 332 10 05/12/22 20220501 NES 220501 1 Introduction Till date, 20 Isotopes of Plutonium (Pu) with mass number 228 to 247 had been discovered. The isotope ratio measurement of Pu isotopes is preferred with thermal ionization mass spectrometer (TIMS) which offers and accuracy and precision better than 0.1% but the technique requires exhaustive sample preparation [[2]]. Application of artificial neural networks for predicting the isotopic composition of high burn-up solid plutonium sample using the 90-105 keV gamma-spectrum region. [Extracted from the article]

Their review work summarizes various casting techniques to fabricate bronze alloys, mainly focusing on the microstructures, tensile properties, and tribological characteristics of Cu-Sn and Cu-Sn-Ti alloys [[19]]. 36676282 2 Li R., Xiao Z., Li Z., Meng X., Wang X. Work Hardening Behavior and Microstructure Evolution of a Cu-Ti-Cr-Mg Alloy during Room Temperature and Cryogenic Rolling. In order to obtain high-quality non-ferrous materials and/or components, we must fully understand the underlying process-structure-property relationship in the metallic materials such as Al alloys, Mg alloys, Ti alloys, Cu alloys, and etc. Huang et al. [[6]] elucidated the precipitation behavior of the T SB B sb (Al SB 7 sb Cu SB 4 sb Li) and S(Al SB 2 sb CuMg) phases of the 2195 Al-Cu-Li alloy at different homogenization temperatures. [Extracted from the article]

(Th, U)O2-based mixed oxide (MOX) fuel is the proposed fuel for advanced heavy water reactors (AHWRs), which belong to the third stage of the three-stage Indian nuclear power programme. The composition of these types of mixed oxide fuels is one of the key parameters in chemical quality control. In the present work, a green analytical methodology has been developed utilizing the micro-X-Ray Fluorescence (micro-XRF) technique for the compositional analysis of these types of fuel pellets. Two different sample preparation methods have been proposed in this work. In Method 1, ≈500–600 μg powder sample scratched from the surface of the pellets was dispersed homogeneously in 1% TRITON solution, and a few hundred nanograms of the sample was deposited on a sample support for micro-XRF analysis. In Method 2, different portions of the pellet made contact with a Scotch tape sample support to transfer a few hundred nanograms of the sample, which was further used for micro-XRF measurements. Both of these methodologies do not require a cumbersome dissolution process in corrosive acids. Moreover, only a few hundred nanograms of the sample amount are sufficient for both methods. Using both sample preparation methodologies, 8 samples in the form of MOX fuel pellets were analyzed, and the results were compared with those obtained from the previously established total reflection X-ray fluorescence (TXRF) technique. The accuracy and precision obtained using Method 1 are 3.06% and 0.99%, respectively, and for Method 2, they are 3.79% and 2.96%, respectively. These methodologies require only a single calibration (or relative sensitivity measurement) for the analysis of all MOX fuel pellets regardless of the composition. [ABSTRACT FROM AUTHOR]

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A preliminary derived concentration guideline level (DCGL) was calculated for the soil of the Wolsong nuclear power plant (NPP) (the first commercial pressurized heavy water reactor in South Korea) site using the RESRAD-ONSITE computational code. In total, fourteen selected radionuclides were analyzed after considering the preliminary evaluation information of radionuclides observed in the pressure tube specimen in the Wolsong Unit 1 heavy water reactor and previous NPP decommissioning cases. Furthermore, a geological structure model of the Wolsong NPP site was established according to the safety analysis report of the Wolsong NPP. In addition, the distribution coefficients (Kd) of various radionuclides were derived using the JAEA-SDB and the pH information of the groundwater around the Wolsong NPP site. The DCGL for surface soil of the Wolsong NPP site was derived via the application of criteria for the site release to facilitate unrestricted reuse. Moreover, preliminary dose evaluation and relevant analysis were performed according to the Wolsong NPP site resident scenario. The novelty of this study lies in the first calculation of the preliminary DCGL values for the case of the pressurized heavy water reactor (Wolsong NPP) site. It is expected that further reliable DCGL results might be achievable if more precise radionuclide information and site-specific parameters with respect to the Wolsong NPP site are secured and applied in the future. [ABSTRACT FROM AUTHOR]

Radiative heat transfer between the different components of the nuclear reactor channel (fuel pins, pressure tube (PT) and calandria tube (CT)) is the predominant mechanism of heat transfer in a channel when coolant flow is ceased. In this paper, the thermal behavior of the channel of Indian PHWR that has thirty-seven fuel elements is analyzed for a postulated accident such as loss of coolant accident (LOCA) coupled with the failure of the emergency core cooling system. (ECCS). The analysis has been performed using experimental, numerical and analytical techniques. The experiment is performed under the pseudo-steady-state condition, and the temperature profiles for the different parts of a simulated channel are obtained. The numerical analysis of the above problem is carried out using the ANSYS® Fluent 19.0. The code calculation is compared with the exact solution which is obtained using the radiosity network method. There is a reasonably good agreement between these three types of analysis. The results from these analyzes show that the temperature gradient (along the radial direction) is developed in the fuel bundle simulator (FBS) from the center fuel element to the outer ring fuel element. Also, an insignificant circumferential temperature gradient is observed in the FBS, PT and CT. [ABSTRACT FROM AUTHOR]