Band 88, Heft 4

In case of a postulated severe accident in a water-cooled nuclear power plant significant amounts of hydrogen (H 2 ) and carbon monoxide (CO) can be generated and released into the containment or reactor building where it might form a combustible mixture with air assuming passive autocatalytic recombiners are not available. In case of ignition, pressure peaks might occur, that are relevant for the integrity of safety relevant equipment and the containment or reactor building. It is therefore important for safety analysis to be able to correctly predict combustion phenomena that might occur. The accident analysis code AC 2 2021.0 which is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) includes the Containment Code System (COCOSYS version 3.1) for the simulation of containment phenomena. COCOSYS contains the model FRONT for the simulation of premixed deflagration of H 2 and CO. Recent code validation using H 2 deflagration tests conducted in the multi-compartment THAI+ test facility shows that the flame propagation stops prematurely in simulations of some tests. This is partly attributed to the missing separation of burned and unburned atmosphere which leads to a reduction in fuel concentration in not yet burning zones connected to a burning zone. Model improvement potential was identified which is addressed in this paper. A model extension to separate burned and unburned atmosphere via a junction model is proposed and implemented into a development version of COCOSYS 3.1. First validation results using the THAI test HD-39 are discussed that show improved prediction capability by the extended model.

The sodium-cooled fast reactor is a Generation-IV International Forum recommended technology, with an aim to improve sustainability, safety, and proliferation resistance. To ensure accurate reactor physics calculation and safety analyses, nuclear data libraries require continuous improvement through modifications based on additional measurements, evaluations, and validation studies with criticality experiments. In this work the Sodium-cooled Fast Reactor Uncertainty Analysis in Modeling (SFR-UAM) benchmark served as a basis to assess differences in nuclear data libraries and estimate variability in criticality and power distribution results. The research has been carried out using the OpenMC code and the study presented here covers two SFR models: MOX-3600 and ABR-1000. The neutronic calculation of numerous parameters in fast spectrum systems including effective multiplication factor ( k eff ), effective delayed neutron fraction ( β eff ), sodium void reactivity (Δ ρ Na ), Doppler constant (Δ ρ Doppler ), and control rod ( ρ CR ) worth were calculated and compared mainly to five libraries: ENDF/B-VII.1, ENDF/B-VIII, JEFF-3.3, JENDL-4.0 and TENDL-2019. In addition, sensitivity calculations using GPT-free method were conducted to understand relevant sensitivities for a given quantity of interest in major isotope/reaction pairs. The major driver of observed uncertainty in k eff are found for the high actinide isotopes mainly capture cross section of 239, 240 Pu as well as fission reaction of 239 Pu.

The liquid radioactive waste generated by the G.A. Siwabessy reactor (RSG-GAS) is categorized into low-activity liquid radwaste (LALR) and medium-activity liquid radwaste (MALR). The radionuclide content of both LALR and MALR can use as an indicator of the structural integrity of the reactor’s systems, structures, and components (SSC). To evaluate the degradation of the reactor SSC, the radionuclide species were identified, and their activities were measured using gamma spectroscopy. Based on the identified radionuclides, the process of their formation can be traced. The radionuclides identified in LALR were 24 Na, 51 Cr, 59 Fe, 60 Co, 65 Zn, and 124 Sb, while the radionuclides in MALR were 24 Na, 51 Cr, 58 Co, 59 Fe, 60 Co, 65 Ni, 65 Zn, 89 Kr, 90 Kr, 109 Cd, 131 I, 132 I, 140 Ba, 137 Cs, 146 Ce, and several others. The radionuclides found can be classified into corrosion product activation ( 60 Co, 65 Zn, 51 Cr, 59 Fe, 24 Na, 65 Ni), topaz impurities activation ( 51 Cr, 59 Fe, 60 Co, 65 Zn), fission product ( 90 Kr, 140 Ba, 131 I, 137 Cs, etc.), and demineralized water impurities activation ( 51 Cr, 59 Fe, 65 Zn, 60 Co, etc.). After comparing the activity value of each radionuclide with the limit value in the safety analysis report document, we can conclude that the activity of each one is below the required level. It can infer that the structural integrity of reactor SSC is still well maintain. During routine monitoring, the radionuclide content in the primary coolant fluctuates depending on the reactor load. The concentration of radionuclides detected varies when a large or small number of research samples are loaded onto the core. Nevertheless, their activities remain within the required safety limits.

The aim of this study is to assess the performance of an existing dry storage cask design for accident tolerant fuel loading case and to examine the compliance with the safety limits applied currently for dry storage. In the study, for a dry storage cask design currently in use, criticality calculations, dose rate evaluation and thermal analyses are performed in case of loading with the accident tolerant fuel discharged from a PWR. Firstly, for an accident tolerant fuel selected among the concepts proposed for use in light water reactors, burnup analyses are performed by utilizing the Serpent continuous energy code and spent fuel characteristics are determined. Then, criticality analyzes are carried out by using the Serpent Monte Carlo code for the case of loading the accident tolerant fuel into the selected dry storage cask design. Gamma and neutron dose rates at the outer surface and close distances of the storage cask are determined with the Serpent code. To evaluate the thermal performance of the storage cask, thermal analyzes are performed by using the ANSYS Fluent computational fluid dynamics code. The analysis results are compared with the nuclear safety criteria applied to dry storage casks. Results of the analysis show that the dry storage cask design currently in-use does not exceed the criticality, dose rate and maximum surface temperature limits when loaded with spent accident tolerant fuel.

Handling the power deposition, reducing erosion effects, and plasma configuration are the key factors in the design of a divertor. The design of Pakistan Spherical Tokamak (PST) is based on double-null divertor configuration with actively cooled graphite targets at outer/inner strike point and peak heat flux range capacity of 0.1–0.3 MW/m 2 . The configuration of PST divertor module is designed with mock-up (used flat type tiles on baffles and dome) and cassette (support PFC and cooling channels) technology. Helium-cooled stage and water-cooled stage are two options for divertor. Therefore, one part of this research is focused on water-cooling system for the divertor. This paper presents the divertor design for PST with cooling channel and material analysis of the divertor, which is carried out in three phases. In the first phase, the plasma edge temperature, density, particle velocity, input power, heat flux, and surface temperature are estimated. In second phase, physics and engineering design of divertor system has been performed. In the third phase, COMSOL simulation has been performed to analyses the material properties, surface temperature rise (∆ T  °C) at stable heat flux, and thermal hydraulic system for the divertor. It is found from the analysis that the specific heat flux of 0.3 MW/m 2 up to 3 s is the safe zone limit. The R & D work ratifies that manufacturing and installation processes are plausible for the proposed divertor design. This design is able to meet the requirement of PST. However, increasing time or specific heat flux beyond these limits would require redesigning of the cooling channel.

In this paper, the time evolution of bremsstrahlung radiation loss, plasma frequency and electron particles density and the relationship between these parameters and black body radiation are investigated. The model used in this work is based on numerical solution of particle and energy balance equations in ITER with DT fuel. The fusion reaction takes places in a plasma of deuterium and tritium heated to millions of degrees. It is expected that at this temperature, the thermal noise could have a significant effect on plasma behavior. This effect is considered in the solution of equations for the first time in this work. In order to attain a proper set of particle and energy balance equations, an appropriate thermal noise term is considered in the set of coupled differential equations. These equations are solved simultaneously by numerical methods. The results of the calculations for bremsstrahlung radiation loss, plasma frequency, intensity of blackbody radiation, absorption coefficient and quality factor show that in the absence of thermal noise blackbody radiation doesn’t occur but in the presence of thermal noise blackbody radiation occurs in times of 55.7 s and 42.73 s for two cases of considering and ignoring impurity respectively. As it can be seen that with the addition of impurities to the system, bremsstrahlung radiation and intensity of blackbody radiation increase while absorption coefficient and quality factor decrease.

In line with population expansion and industrial development, the world’s energy consumption has been rising gradually over the past three decades. As a result, methods for energy conservation have been sought. One of the most common strategies is heat recovery, which is efficient and cost-effective to the extent possible. Heat recovery is not just about saving energy for primary consumption; it is also about lowering emissions and protecting the environment. In this respect, one of the most important strategies for heat recovery is to develop heat exchangers and exploit the energy associated with many of the processes’ output products in order to use it in new processes. Many researchers working in the field of heat engineering are now looking into novel heat transfer techniques. Use of the heat exchanger as a compact is one of these ways that might be considered. The current review therefore concentrates on the design of plate-fin heat exchangers (PFHE) and multi-stream plate-fin heat exchangers (MSPFHE) based on various models. The current review offers some suggestions for upcoming studies on improving heat transfer and minimizing power use.

Many countries are considering Small and Modular Reactors as a viable alternative to counter the climate-change/global-warming with a quick deployment of green, carbon free nuclear energy option in the energy mix. Proponents of SMRs claim that these designs rely more on enhanced inherent/engineered safety and passive features with novel concepts. SMRs are being designed to be fabricated at a factory and then transported as ‘modules’ to the sites for installation either as a single module or multiple module plant. There are many variant of SMRs under considerations/design/construction/commissioning/operation states and majority of the, more than 70 odd SMRs are in the design stage. The paper focuses on safety aspects while addressing the fundamental safety requirement that are derived from fundamental safety principles, the acceptance criteria, the expected/envisaged safety targets and not only the economic impact/considerations. The assessment basis for requirements towards safety enhancements and their extent of assurance in the design are highlighted against the claims made. Ensuring SMR safety with respect to the fundamental safety functions will depend on the foreseen/predicted fission product releases, following overheating of the fuel, during the worst/credible accident conditions and likelihood of occurrence of these accidents. Innovations in the development of advanced fuel, deploying passive safety systems, novel concepts in main heat transport system configuration and advanced features in instrumentation can help in realising the goal of ensured enhanced safety in the SMRs, both in preventive and mitigation domains during severe accidents. Enhancements in the acceptance criteria and deterministic and probabilistic safety targets is also expected and may be envisaged. The paper brings out the challenges faced in the design and regulation of the new NPPs, while addressing fundamental safety principles implementation, generic, specific safety issues, and only genuine innovations can ensure and improve the safety. Aspects related to passive systems and the optimal main heat removal system configuration of the NPPs are also discussed. The aspects related to concurrent design and regulation of new NPPs including SMRs also has been brought out in the paper.

Study on calculation model and risk area of radionuclide diffusion in coastal waters under nuclear leakage accidents with different levels can help predict and evaluate consequences of radionuclide leakage accidents. Thus they play an important role in emergency response and accident mitigation. In the first step of the study, a climate hydrodynamic model in China coastal waters was established based on the climate data. In the next step, according to the real-time meteorological data, a hydrodynamic model in coastal waters of Haiyang nuclear power station was founded using the result of the climate hydrodynamic as a boundary. Then, according to the result of the hydrodynamic model in coastal waters of Haiyang nuclear power station, a radionuclide diffusion model in coastal waters of Haiyang nuclear power station was set up, in which the Euler method was adopted. With the radionuclide diffusion model, the total leaked radioactivity of radionuclides was set from 10 18  Bq to 10 12  Bq with a decrease of every two orders of magnitude. Thus, scenarios of radionuclide diffusion under assumed nuclear leakage accidents with different levels were calculated and their corresponding risk area were analyzed under the assumption that radionuclides leaked for consecutive five days. The results show that when the leaked radioactivity of radionuclides is 10 18  Bq, the risk area on the seventh day is about 41 km east, 22 km south and 19 km west of the power station; on the fourteenth day, the risk area is about 65 km east, 22 km south and 25 km west of the power station. When the total leaked radioactivity of radionuclides declines by two orders of magnitude, the risk area will be reduced by about 10 km–20 km in the east direction accordingly. When it declines to 10 14  Bq, the risk area decreases sharply to a small area. When it declines to 10 12  Bq, the risk area is barely found. This model was verified from two aspects, namely the flow field and the radionuclide concentration. Hydrodynamic results can well describe the Yellow Sea cold water mass, Yellow Sea warm current and tidal current. Changes of radioactivity in different positions are fundamentally consistent with that in Fuikushima nuclear leakage accident. It indicates the hydrodynamic model and radionuclide diffusion model in the study are feasible and reliable.

TRISO fuel particle using ZrC has better strength and resistance to high temperatures than SiC. Previous studies show that the ZrC layer, as a substitution of SiC within the TRISO layer of coated fuel particles, has an insignificant difference in the performance of the neutronic aspect. Further neutronic studies are required to obtain the best combination of thorium-based fuel with ZrC coating for HTGR. This study analyzed the neutronic performance of three types of thorium-based fuels, oxide, carbide, and nitride, for HTGR. The reactor design refers to the High-Temperature Test Reactor with some axial and radial fuel configuration adjustments. This reactor is designed to operate at 200 MWt and has been modified to use a ZrC layer as a substitute for the SiC layer on the coated fuel particles. The neutronic study is carried out using SRAC2006 code with JENDL 4.0 nuclear data library. Neutronic parameters analyzed include multiplication factor, power peaking factor, and neutron spectrum. Neutronic analysis results show that thorium nitride fuel’s multiplication factor ( k eff ) is better than other compared fuel types with k -eff 1.050, higher than thorium carbide, 1.004. At the same time, thorium oxide has been sub-critical. The power-peaking value of all materials is close to the ideal peaking value that is one. Other neutronic aspects, such as the neutron spectrum for three compared fuel types, have a similar trend.

This investigation reports on the experimental outcomes of the pool boiling heat transfer characteristics, specifically on the downward heated surface, concerning reverse osmosis water and γ -Fe 2 O 3 nanofluids. To conduct the pool boiling experiments, γ -Fe 2 O 3 nanofluids were prepared with variable concentrations ranging from 2 mg/L to 10 mg/L. Analysis of the experimental data revealed that a concentration of 5 mg/L yielded the greatest enhancement effect on critical heat flux (CHF), with an increase of 13.5 %. However, the results also indicated that excessively high concentrations of nanofluid had a negative impact on CHF enhancement. The impact of nanofluids on heat transfer performance was investigated by analyzing the observed bubble behavior during the boiling process, measuring the drop angle and surface roughness post-experiment, and characterizing the heated surface morphology via scanning electron microscopy (SEM). Through these methods, the underlying mechanism behind the impact of nanofluids on heat transfer performance was identified and analyzed.

The radial non-uniformity of an ADS fuel rod power density on the peak plane is as high as 10.46 %, which is derived from the reactor physics calculations. In order to investigate the influence of the above-mentioned radial power inhomogeneity on the fuel temperature distribution, this paper constructs a set of two-dimensional comparison examples, where one example uses a uniform heat source and another uses a non-uniform heat source distribution, with taking the fuel segment and its corresponding cladding segment in the region where the peak plane is located as the research object. The finite element software COMSOL is used to conduct the heat transfer analysis. The results of the study show that the fuel temperature under radial non-uniform power distribution is almost the same as that under uniform power. Therefore, this radial non-uniformity can be completely ignored when the research object of temperature is considered. The quantitative calculation carried out in this research can provide certain data support for the engineering research of accelerator driven subcritical system, and can also provide certain guidance for the performance analysis of such fuel elements.

Heat transfer enhancement is required for numerous situations, i.e., gas turbines, nuclear power plants, micro and macro scale heat transfer, airfoil cooling, electronic cooling, semiconductors, biomedical and combustion chamber lines, etc. One of the prominent ways of increasing the heat transfer coefficient from the surface of a heat exchanger is by moving the position of the thermal boundary layer to make it either thinner or break the same partially. It requires making use of an increased surface area/fins. Accordingly, the research progressed in heat transfer enhancement by using concavities/dimples of the heat exchanger surfaces to improve the heat transfer coefficient and heat transfer rate. These impregnations are made on the internal flow tubes/surfaces of the heat exchanger surfaces. The present research work aims at the experimental investigation of a heat exchanger to determine the airflow pattern and computation of heat transfer rate on the dimpled surfaces. This research work will be beneficial and applicable to heat transfer enhancement applications like micro heat transfer, where space constraint is considered. The geometries considered for the experiment include flat plates and dimpled surfaces. The parameters like Reynolds number (varied from 20,000 to 50,000), dimple depth to imprint diameter ratio (varied from 0.2 to 0.4), and heater input to the test plates (varied from 75 to 120 W) are considered for the comparisons. The results with dimpled surfaces are compared with the flat plate surfaces having no dimples. The Reynolds and Nusselt numbers rise in direct proportion to the heater input. For pin fin and dimpled plate, the ratio of Nusselt number to area average Nusselt number drops for 75 W and 100 W input. The dimpled plate with a ratio of 0.3 between imprint diameter to dimple depth had the highest ratio of Nusselt number to Nusselt number value for the entire group.