Abstract: In this study, we focus on the establishment of the analysis model for Maanshan PWR nuclear power plant (NPP) by using RADTRAD and SNAP codes with the FSAR, manuals, and other data. In order to evaluate the cumulative dose at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) outer boundary, Maanshan NPP RADTRAD/SNAP model was used to perform the analysis of the DBA LOCA case. The analysis results of RADTRAD were similar to FSAR data. These analysis results were lower than the failure criteria of 10 CFR 100.11 (a total radiation dose to the whole body, 250 mSv; a total radiation dose to the thyroid from iodine exposure, 3000 mSv).
Abstract: In this research, the HABIT analysis methodology was established for Maanshan nuclear power plant (NPP). The Final Safety Analysis Report (FSAR), reports, and other data were used in this study. To evaluate the control room habitability under the CO2 storage burst, the HABIT methodology was used to perform this analysis. The HABIT result was below the R.G. 1.78 failure criteria. This indicates that Maanshan NPP habitability can be maintained. Additionally, the sensitivity study of the parameters (wind speed, atmospheric stability classification, air temperature, and control room intake flow rate) was also performed in this research.
Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of Spent Fuel Pools (SFPs) in Taiwan after Fukushima event. In order to estimate the safety of Kuosheng NPP SFP, by using MELCOR2.1 and SNAP, the safety analysis of Kuosheng NPP SFP was performed combined with the mitigation strategy of NEI 06-12 report. There were several steps in this research. First, the Kuosheng NPP SFP models were established by MELCOR2.1/SNAP. Second, the Station Blackout (SBO) analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition. The results showed that the calculations of MELCOR and TRACE were very similar in this case. Second, the mitigation strategy analysis was done with the MELCOR model by following the NEI 06-12 report. The results showed the effectiveness of NEI 06-12 strategy in Kuosheng NPP SFP. Finally, a sensitivity study of SFP quenching was done to check the differences of different water injection time and the phenomena during the quenching. The results showed that if the cladding temperature was over 1600 K, the water injection may have chance to cause the accident more severe with more hydrogen generation. It was because of the oxidation heat and the “Breakaway” effect of the zirconium-water reaction. An animation model built by SNAP was also shown in this study.
Abstract: In this study of ultimate response guideline (URG), Kuosheng BWR/6 nuclear power plant (NPP) TRACE model was established. The reactor depressurization, low pressure water injection, and containment venting are the main actions of URG. This research focuses to evaluate the efficiency of URG under Fukushima-like conditions. Additionally, the sensitivity study of URG was also performed in this research. The analysis results of TRACE present that URG can keep the peak cladding temperature (PCT) below 1088.7 K (the failure criteria) under Fukushima-like conditions. It implied that Kuosheng NPP was at the safe situation.
Abstract: The dismantling of disused industrial facilities such as nuclear power plants or refineries is an enormous challenge for the planning and control of the logistic processes. Existing control models do not meet the requirements for a proper dismantling of industrial plants. Therefore, the paper presents an approach for the control of dismantling and post-processing processes (e.g. decontamination) in plant decommissioning. In contrast to existing approaches, the dismantling sequence and depth are selected depending on the capacity utilization of required post-processing processes by also considering individual characteristics of respective dismantling tasks (e.g. decontamination success rate, uncertainties regarding the process times). The results can be used in the dismantling of industrial plants (e.g. nuclear power plants) to reduce dismantling time and costs by avoiding bottlenecks such as capacity constraints.
Abstract: Nuclear technology is a controversial issue among a great share of the Brazilian population. Misinformation and common wrong beliefs confuse public’s perceptions and the scientific community is expected to offer a wider perspective on the benefits and risks resulting from ionizing radiation in everyday life. Attentive to the need of new approaches between science and society, the Nuclear Energy Museum, in northeast Brazil, is an initiative created to communicate the growing impact of the beneficial applications of nuclear technology in medicine, industry, agriculture and electric power generation. Providing accessible scientific information, the museum offers a rich learning environment, making use of different educational strategies, such as films, interactive panels and multimedia learning tools, which not only increase the enjoyment of visitors, but also maximize their learning potential. Developed according to modern active learning instructional strategies, multimedia materials are designed to present the increasingly role of nuclear science in modern life, transforming science education into a meaningful learning experience. In year 2016, nine different interactive computer-based activities were developed, presenting curiosities about ionizing radiation in different landmarks around the world, such as radiocarbon dating works in Egypt, nuclear power generation in France and X-radiography of famous paintings in Italy. Feedback surveys have reported a high level of visitors’ satisfaction, proving the high quality experience in learning nuclear science at the museum. The Nuclear Energy Museum is the first and, up to the present time, the only permanent museum in Brazil devoted entirely to nuclear science.
Abstract: After the measurement uncertainty recapture (MUR) power uprates, Kuosheng nuclear power plant (NPP) was uprated the power from 2894 MWt to 2943 MWt. For power upgrade, several codes (e.g., TRACE, RELAP5, etc.) were applied to assess the safety of Kuosheng NPP. Hence, the main work of this research is to establish a RELAP5/MOD3.3 model of Kuosheng NPP with SNAP interface. The establishment of RELAP5/SNAP model was referred to the FSAR, training documents, and TRACE model which has been developed and verified before. After completing the model establishment, the startup test scenarios would be applied to the RELAP5/SNAP model. With comparing the startup test data and TRACE analysis results, the applicability of RELAP5/SNAP model would be assessed.
Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of NPPs in Taiwan after Japan Fukushima NPP disaster occurred. Hence, in order to estimate the safety of Kuosheng NPP spent fuel pool (SFP), by using TRACE, MELCOR, and SNAP codes, the safety analysis of Kuosheng NPP SFP was performed. There were two main steps in this research. First, the Kuosheng NPP SFP models were established. Second, the transient analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition (Fukushima-like condition). The results showed that the calculations of MELCOR and TRACE were very similar in this case, and the fuel uncover happened roughly at 4th day after the failure of cooling system. The above results indicated that Kuosheng NPP SFP may be unsafe in the case of long-term SBO situation. In addition, future calculations were needed to be done by the other codes like FRAPTRAN for the cladding calculations.
Abstract: The decreasing use of fossil fuel power stations has
a negative effect on the stability of the electricity systems in many
countries. Nuclear power stations have traditionally provided minimal
ancillary services to support the system but this must change in the
future as they replace fossil fuel generators. This paper explains the
development of the four most popular reactor types still in regular
operation across the world which have formed the basis for most
reactor development since their commercialisation in the 1950s. The
use of nuclear power in four countries with varying levels of capacity
provided by nuclear generators is investigated, using the primary
frequency response provided by generators as a measure for the
electricity networks stability, to assess the need for nuclear generators
to provide additional support as their share of the generation capacity
increases.
Abstract: Natural circulation loops (NCLs) are buoyancy driven flow systems without any moving components. NCLs have vast applications in geothermal, solar and nuclear power industry where reliability and safety are of foremost concern. Due to certain favorable thermophysical properties, especially near supercritical regions, carbon dioxide can be considered as an ideal loop fluid in many applications. In the present work, a high temperature NCL that uses supercritical carbon dioxide as loop fluid is analysed. The effects of relevant design and operating variables on loop performance are studied. The system operating under steady state is modelled taking into account the axial conduction through loop fluid and loop wall, and heat transfer with surroundings. The heat source is considered to be a heater with controlled heat flux and heat sink is modelled as an end heat exchanger with water as the external cold fluid. The governing equations for mass, momentum and energy conservation are normalized and are solved numerically using finite volume method. Results are obtained for a loop pressure of 90 bar with the power input varying from 0.5 kW to 6.0 kW. The numerical results are validated against the experimental results reported in the literature in terms of the modified Grashof number (Grm) and Reynolds number (Re). Based on the results, buoyancy and friction dominated regions are identified for a given loop. Parametric analysis has been done to show the effect of loop diameter, loop height, ambient temperature and insulation. The results show that for the high temperature loop, heat loss to surroundings affects the loop performance significantly. Hence this conjugate heat transfer between the loop and surroundings has to be considered in the analysis of high temperature NCLs.
Abstract: In France, in the main media, the concern about nuclear safety and security has not really appeared before the beginning of the 1970s. The gradual changes in its perception are studied here through the arguments given in the main French news magazines, linked with several parameters. As this represents a considerable amount of copies and thus of information, are selected here the main articles as well as the main “mental images” aiming to persuade the readers and which have led the public awareness to evolve. Indeed, in the 1970s, in France, these evolutions were not made in one day. Indeed, over the period, many articles were still in favor of nuclear power plants and promoted the technological advances that were made in this field. They had to be taken into account. But, gradually, grew up arguments and mental images discrediting the perception of nuclear technology. Among these were the environmental impacts of this industry, as the question of pollution progressively appeared. So, between 1970 and 1979, the language has changed, as the perceptible objectives of the communication, allowing to discern the deepest intentions of the editorial staffs of the French news magazines. This is all these changes that are emphasized here, over a period when the safety and security concern linked to the nuclear technology, to there a field for specialists, has become progressively a social issue seemingly open to all.
Abstract: The performances of nuclear fuels and materials are qualified at an irradiation system in research reactors operating under the commercial nuclear power plant conditions. Fuel centerline temperature, coolant temperature, neutron flux, deformations of fuel stack and swelling are important parameters needed to analyze the nuclear fuel performances. The dimensional stability of nuclear fuels is a key parameter measuring the fuel densification and swelling. In this study, the fuel stack elongation is measured using a LVDT. A mockup LVDT instrumented fuel rod is developed. The performances of mockup LVDT instrumented fuel rod is evaluated by experiments.
Abstract: TRACE is developed by U.S. NRC for the nuclear
power plants (NPPs) safety analysis. We focus on the establishment
and application of TRACE/FRAPTRAN/SNAP models for Chinshan
NPP (BWR/4) spent fuel pool in this research. The geometry is 12.17
m × 7.87 m × 11.61 m for the spent fuel pool. In this study, there are
three TRACE/SNAP models: one-channel, two-channel, and
multi-channel TRACE/SNAP model. Additionally, the cooling system
failure of the spent fuel pool was simulated and analyzed by using the
above models. According to the analysis results, the peak cladding
temperature response was more accurate in the multi-channel
TRACE/SNAP model. The results depicted that the uncovered of the
fuels occurred at 2.7 day after the cooling system failed. In order to
estimate the detailed fuel rods performance, FRAPTRAN code was
used in this research. According to the results of FRAPTRAN, the
highest cladding temperature located on the node 21 of the fuel rod
(the highest node at node 23) and the cladding burst roughly after 3.7
day.
Abstract: As the nuclear power is a sensitive field leading to controversy, the quality of the communication about it is important. Between 1965 and 1981, in France, this one had gradually changed. This change is studied here in the main French news magazine L’Express, in connection with several parameters. As this represents a huge number of copies and occurrences, thus a considerable amount of information; this paper is focused on the main articles as well as the main “mental images”. These ones are important, as their aim is to direct the thought of the readers, and as they have led the public awareness to evolve. Over this 17 years, two trends are in confrontation: The first one is promoting the perception of the nuclear power, while the second one is discrediting it. These trends are organized in two axes: the evolution of engineering, and the risks. In both cases, the changes in the language allow discerning the deepest intentions of the magazine editing, over a period when the nuclear technology, to there a laboratory object accompanied with mystery and secret, has become a social issue seemingly open to all.
Abstract: Ensuring of continuity of business is basic strategy of
every company. Continuity of organization activities includes
comprehensive procedures that help in solving unexpected situations
of natural and anthropogenic character (for example flood, blaze,
economic situations). Planning of continuity operations is a process
that helps identify critical processes and implement plans for the
security and recovery of key processes. The aim of this article is to
demonstrate application of system approach to managing business
continuity called business continuity management systems in military
issues. This article describes the life cycle of business continuity
management which is based on the established cycle PDCA (Plan-
Do-Check-Act). After this is carried out by activities which are
making by University of Defence during activation of forces and
means of the integrated rescue system in case of emergencies -
accidents at a nuclear power plant in Czech Republic. Activities of
various stages of deployment earmarked forces and resources are
managed and evaluated by using MCMS application (Military
Continuity Management System).
Abstract: A Multi-dimensional computational fluid dynamics
(CFD) two-phase model was developed with the aim to simulate
the in-core coolant circuit of a pressurized heavy water reactor
(PHWR) of a commercial nuclear power plant (NPP). Due to the
fact that this PHWR is a Reactor Pressure Vessel type (RPV),
three-dimensional (3D) detailed modelling of the large reservoirs of
the RPV (the upper and lower plenums and the downcomer) were
coupled with an in-house finite volume one-dimensional (1D) code
in order to model the 451 coolant channels housing the nuclear fuel.
Regarding the 1D code, suitable empirical correlations for taking into
account the in-channel distributed (friction losses) and concentrated
(spacer grids, inlet and outlet throttles) pressure losses were used.
A local power distribution at each one of the coolant channels
was also taken into account. The heat transfer between the coolant
and the surrounding moderator was accurately calculated using a
two-dimensional theoretical model. The implementation of subcooled
boiling and condensation models in the 1D code along with the use
of functions for representing the thermal and dynamic properties of
the coolant and moderator (heavy water) allow to have estimations
of the in-core steam generation under nominal flow conditions for a
generic fission power distribution. The in-core mass flow distribution
results for steady state nominal conditions are in agreement with the
expected from design, thus getting a first assessment of the coupled
1/3D model. Results for nominal condition were compared with
those obtained with a previous 1/3D single-phase model getting more
realistic temperature patterns, also allowing visualize low values of
void fraction inside the upper plenum. It must be mentioned that the
current results were obtained by imposing prescribed fission power
functions from literature. Therefore, results are showed with the aim
of point out the potentiality of the developed model.
Abstract: Growing human population has placed increased
demands on water supplies and spurred a heightened interest in
desalination infrastructure. Key elements of the economics of
desalination projects are thermal and electrical inputs. With growing
concerns over use of fossil fuels to (indirectly) supply these inputs,
coupling of desalination with nuclear power production represents a
significant opportunity. Individually, nuclear and desalination
technologies have a long history and are relatively mature. For
desalination, Reverse Osmosis (RO) has the lowest energy inputs.
However, the economically driven output quality of the water
produced using RO, which uses only electrical inputs, is lower than the
output water quality from thermal desalination plants. Therefore,
modern desalination projects consider that RO should be coupled with
thermal desalination technologies (MSF, MED, or MED-TVC) with
attendant steam inputs to permit blending to produce various qualities
of water. A large nuclear facility is well positioned to dispatch large
quantities of both electrical and thermal power. This paper considers
the supply of thermal energy to a large desalination facility to examine
heat balance impact on the nuclear steam cycle. The APR1400 nuclear
plant is selected as prototypical from both a capacity and turbine cycle
heat balance perspective to examine steam supply and the impact on
electrical output. Extraction points and quantities of steam are
considered parametrically along with various types of thermal
desalination technologies to form the basis for further evaluations of
economically optimal approaches to the interface of nuclear power
production with desalination projects. In our study, the
thermodynamic evaluation will be executed by DE-TOP, an IAEA
sponsored program. DE-TOP has capabilities to analyze power
generation systems coupled to desalination plants through various
steam extraction positions, taking into consideration the isolation loop
between the nuclear and the thermal desalination facilities (i.e., for
radiological isolation).
Abstract: In this research, TRACE model of Chinshan BWR/4
nuclear power plant (NPP) has been developed for the simulation and
analysis of ultimate response guideline (URG).The main actions of
URG are the depressurization and low pressure water injection of
reactor and containment venting. This research focuses to verify the
URG efficiency under Fukushima-like conditions. TRACE analysis
results show that the URG can keep the PCT below the criteria
1088.7 K under Fukushima-like conditions. It indicated that Chinshan
NPP was safe.
Abstract: Reflux condensation occurs in vertical channels and tubes when there is an upward core flow of vapour (or gas-vapour mixture) and a downward flow of the liquid film. The understanding of this condensation configuration is crucial in the design of reflux condensers, distillation columns, and in loss-of-coolant safety analyses in nuclear power plant steam generators. The unique feature of this flow is the upward flow of the vapour-gas mixture (or pure vapour) that retards the liquid flow via shear at the liquid-mixture interface. The present model solves the full, elliptic governing equations in both the film and the gas-vapour core flow. The computational mesh is non-orthogonal and adapts dynamically the phase interface, thus produces a sharp and accurate interface. Shear forces and heat and mass transfer at the interface are accounted for fundamentally. This modeling is a big step ahead of current capabilities by removing the limitations of previous reflux condensation models which inherently cannot account for the detailed local balances of shear, mass, and heat transfer at the interface. Discretisation has been done based on finite volume method and co-located variable storage scheme. An in-house computer code was developed to implement the numerical solution scheme. Detailed results are presented for laminar reflux condensation from steam-air mixtures flowing in vertical parallel plate channels. The results include velocity and gas mass fraction profiles, as well as axial variations of film thickness.
Abstract: In regards to the energy sector in the modern period,
two points were raised. First is a vast and growing energy demand, and
second is an environmental impact associated with it. The enormous
consumption of fossil fuel to the mobile unit is leading to its rapid
depletion. Nuclear power is not the only problem. A modal shift that
utilizes personal transporters and independent power, in order to
realize a sustainable society, is very effective. The author proposes that
the world will continue to work on this. Energy of the future society,
innovation in battery technology and the use of natural energy is a big
key. And it is also necessary in order to save on energy consumption.