Top PDF Model of an Accident-Induced Fire around a Nuclear Power Plant

Model of an Accident-Induced Fire around a Nuclear Power Plant

Model of an Accident-Induced Fire around a Nuclear Power Plant

metodami. PDE opisujejo fizikalne postopke in se rešujejo v trorazsežnem prostoru. Ker požarni modeli štejejo veliko fizikalnih pojavov, postane interakcija med njimi zelo zahtevna in zahteva uporabo raèunalnika. Opisani model požara predstavlja obnašanje požara na odprtem oziroma razvoj dimnega oblaka. Za simulacijo je bil uporabljen program FDS, ki je zasnovan na turbulentnem modelu velikih vrtincev. Disipacijski pojavi se v modelu raèunajo z (niè enaèbenim) modelom Smagorinskega. Prispevek prikazuje, da je natanènost rezultatov zelo odvisna od pravilne predpostavke zaèetnih in robnih pogojev in zadostne gostote numeriène mreže. Zaradi razmeroma velike geometrijske oblike se je po veè simulacijah izkazala potreba po zmanjšanju stalnice Smagorinskega, kakor je bila privzeta v programu FDS.
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RESISTANCE OF NUCLEAR POWER PLANT STRUCTURES TO A LARGE AIRPLANE CRASH AND FIRE

RESISTANCE OF NUCLEAR POWER PLANT STRUCTURES TO A LARGE AIRPLANE CRASH AND FIRE

Sandia National Laboratories in the USA performed the missile crash tests in order to evaluate the dynamic behavior of the fluid according to the high speed impact. Among the related researches, one of the test results is known to provide useful data to evaluate the dispersion of the fluid. In this study, a dynamic model of the fluid is proposed and verified through comparison with the study results of Jepsen et al. (2004) and Brown et al. (2012). The dynamic behavior of the fluid is idealized by SPH method that discretizes the liquid to Lagrange sphere. In this study, the liquid discrete model using SPH method is also applied to the dynamic analysis of the fluid to evaluate dispersing range of the liquid after a fluid- filled cylinder missile crashes to a target structure.
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Accuracy of the grid size for fire modelling in nuclear power plant scenarios

Accuracy of the grid size for fire modelling in nuclear power plant scenarios

Figure 20 shows a dispersion between the heat flux measured in “N U-4” position. Similar results can be observed in cases 1, 2 and 5 (lower cell sizes in the pool fire area), and greater dispersion in the other cases. Figure 21 shows the higher dispersion for the cases 4 and 3. In the case of the measurement point “C C-6”, a larger dispersion is observed in the results obtained for the case 5 compared with the cases 1 and 2. This is because, in case 5, point “C C-6” is located in the simulation area where the grid cell size used was 20 cm instead of 10 cm. Heat flux obtained in the case 4 is lower than the heat flux obtained in the other simulations. It can be observed in the Figure 23 that the heat flux loss by the door is similar in all cases with the exception of the case 4, in which more heat is lost from the compartment by radiation. It can therefore be concluded that the FDS model is very sensitive to the radiation heat transfer.
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Simulation of Nuclear Power Plant Containment Response During a Large-Break Loss-of-Coolant Accident

Simulation of Nuclear Power Plant Containment Response During a Large-Break Loss-of-Coolant Accident

CONTAIN is a lumped-parameter code, which treats a containment system as a network of interconnected control volumes or “cells”. In each cell, fluids are stagnant and homogeneous. The cells represent an actual internal containment compartment or group of compartments. In some cases, a com- partment may be partitioned into several cells to model phenomena such as natural convection or gas stratification. The cells communicate with each other by means of the mass flow of material or heat con- duction through intermediate heat-transfer structures. CONTAIN is designed to use a relatively small num- ber of cells.
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Modeling Migration of Cs-137 in Sewer System of Fukushima City Using Model for Radionuclide Migration in Urban Environment and Drainage System (MUD)

Modeling Migration of Cs-137 in Sewer System of Fukushima City Using Model for Radionuclide Migration in Urban Environment and Drainage System (MUD)

Fig. 4. Monthly average of suspended solid removal of primary sedimentation and secondary clarifier in WWTP Fukushima city [11] Following Fukushima Dai-Ichi Nuclear Power Plant accident, the WWTP has to deal with the high amount of radiocesium content in its sludge. Table 3 shows empirical data of monthly average of activity concentration for Cs-137 and Cs-134 in dewatered sludge. It is shown that activity concentration of Cs-137 and Cs-134 was relatively high during June to September period compared to other months. It is expected that seasonal variation factor plays as the important factor on fluctuation of Cs-137 and Cs-134 activity concentration since the trend has high correlation with rainfall depth data.
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Component Fragility for use in PSA of Nuclear Power Plant

Component Fragility for use in PSA of Nuclear Power Plant

The wind PSA systems model includes wind-caused initiating events and other failures that can lead to core damage or large early release. Typically, the wind PSA systems model is adapted from the internal events PSA systems model by incorporating the wind-analysis aspects that are different from the corresponding aspects in the internal events PSA systems model. Further, the analysis consists of developing event trees and fault trees in which the initiating event can be either the extreme wind effect itself or a transient induced by the extreme winds. Various accident sequences that lead to core damage (Level 1 PSA) or releases (Level 2 PSA) are identified, and their conditional probabilities of occurrence are calculated. The frequency of core damage or release is obtained by a convolution over the relevant range of wind speeds. Factors to be considered in this analysis include specific wind fragilities of SSCs and non-wind-related unavailability or failures of equipment, operator error, any warning time available to take mitigating steps (e.g., in the case of hurricanes), the possibility of recovery actions by operators and replacements by substitutes to accomplish the needed function, and the likelihood of common-cause failures. Since the wind hazard and wind fragilities are input as families of curves representing the uncertainties, the quantification should properly propagate the uncertainties through the accident sequences to result in uncertainty bounds on CDF and release frequencies such as LERF.
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Study on severe accident management measures for the pinping system in nuclear power plant using vibration control techniques

Study on severe accident management measures for the pinping system in nuclear power plant using vibration control techniques

After the accident in Fukushima, it becomes more a mandatory action to enhance the safety level of the nuclear power plant. It is highly requested to suggest measures to severe accident when input motion level exceeds the design basis earthquake. Most of reactions to severe accident occurred at the plant is enhancement of structural internal force. However, there is a limit to heighten the strength of the structure in order to enhance the seismic safety. As metal material has elastic and plastic deformation ranges, it is known that the latter shows wider deformation range than the former. All mechanical components are generally designed to behaviour in elastic range for function maintenance while some mechanical components are fine to reflect elasto-plastic elements. Of most well-known is piping support element. It increases safety of the piping as piping support element deforms plastically by absorbing the energy when responding to the earthquake. It is considered one of the best-possible measures to secure the safety level in the area beyond the design basic accident such as severe accident measures. This study reviewed the vibration control capacity of the piping system by using the elasto-plastic damper. This paper compares the analytical model of the coil spring damper with the mechanical characteristics obtained from the experiments. Test result agrees well with the basic mechanical characteristic of the elasto-plastic damper. This paper discusses on the challenge of thermal expansion from pipe support area and their responses during earthquake.
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Practical Application   of Restraint of Pressure-induced Bending Phenomenon in Leak Rate Calculations   (G412)

Practical Application of Restraint of Pressure-induced Bending Phenomenon in Leak Rate Calculations (G412)

To assess the magnitude of the effect of restraint of pressure-induced bending on real plant piping, a finite element model of a 3-loop Westinghouse-style PWR nuclear power plant was developed, and hinges were placed at eighteen critical locations per the procedure given above. Figures 7 through 10 show the 18 locations, all of which were at high stress points or at field welds. Table 3 provides the pertinent dimensional and loading conditions for these particular locations. Because the angular position of the postulated leaking flaw was not known, the stiffness was calculated at 15-degree intervals around the pipe circumference at each location. The worst stiffness was assumed to correspond with the orientation of the flaw, and this stiffness was used in subsequent calculations.
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Fire safety simulation of cable fire in nuclear power plant room based on flammability database of cables

Fire safety simulation of cable fire in nuclear power plant room based on flammability database of cables

In the numerical simulation, an electrical switchgear rooms in nuclear plants were applied as an evaluation room model. Fire in motor control cabinet in the switchgear room on nearby cable tray was applied to the fire source. From numerical results, the heat configuration and temperature in the switchgear room was evaluated and radiation heat from fire to the cable trays was calculated.

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Assessment of Dispersion Characteristics and Radiation Doses Consequences of a postulated Accident at a Proposed Nuclear Power Plant

Assessment of Dispersion Characteristics and Radiation Doses Consequences of a postulated Accident at a Proposed Nuclear Power Plant

consequences. RASCAL code[5] which stands for Radiological Assessment System for Consequence Analysis, is the software developed and used by the U. S. Nuclear Regulatory Commission (NRC), Emergency Operations Center in order to estimate the projected doses in case of radiological emergencies. The mitigation actions of the emergency plan should take place in the first few hours after an accidental release of radioactivity to atmosphere; model predictions will supplement monitoring data to increase understanding of the radiological situation and to form a basis for emergency health protection decisions.The nuclear power plants emergency plans include preparations for evacuation, sheltering, or other actions to protect the residents near nuclear power plants in the event of a serious incident. Each plant operator is required to exercise its emergency plan with offsite authorities at least once every two years to ensure state and local officials remain proficient in implementing their emergency plans.
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Simulation of a Hypothetical Loss-of-Feedwater Accident in a Modernized Nuclear Power Plant

Simulation of a Hypothetical Loss-of-Feedwater Accident in a Modernized Nuclear Power Plant

Kakor je bilo že reèeno, se je primarni tlak zaèel zviševati (sl. 5), zato so najprej ugasnili grelniki v tlaèniku, ki med obièajnim obratovanjem uravnavajo primarni tlak. Primarni tlak se je namreè hitro zvišal prek meja, med katerimi je predvideno krmiljenje z grelniki. Takoj zatem so se odprli ventili prh in v parni prostor tlaènika je zaèelo pršeti primarno hladivo iz hladnih vej. Ker tudi to ni ustavilo zvišanja primarnega tlaka, sta se eden za drugim odprla oba razbremenilna ventila na tlaèniku. Tako se je zaèelo izgubljati primarno hladivo in celoten primarni sistem je kmalu prešel v stanje nasièenja. Primarni tlak in temperatura sta se tedaj ustalila. Zaradi izgube podhladitve primarnega hladiva (sl. 6) je moral operater v skladu z obratovalnimi navodili za ukrepanje v sili (EOP) zaustaviti obe èrpalki primarnega hladiva. Èrpalki namreè ob delovanju s skupno moèjo 6 MW dodajata toplotno energijo primarnemu hladivu. Potek krivulj, ki prikazujejo podhladitev primarnega hladiva je v obeh simulacijah podoben, nekaj veè razlike je opaziti le v zaèetni vrednosti. Temperatura na izhodu iz sredice se v modelu za RELAP5 meri le v vozlišèu neposredno nad sredico in je pravzaprav povpreèna temperatura hladiva nad sredico. Vhodni model v popolnem simulatorju JEK pa vsebuje 3D model sredice, v katerem se temperatura na izhodu iz sredice
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Rainfall erosivity in catchments contaminated with fallout from the  Fukushima Daiichi nuclear power plant accident

Rainfall erosivity in catchments contaminated with fallout from the Fukushima Daiichi nuclear power plant accident

Based on monthly and annual mean values for each of the 42 stations, 13 GAMs were fitted to spatially model the R factor over the study area (i.e. one model per month and one model for the year). A Gaussian distribution model in- corporated the conditional mean µ (Y ) and, due to the pre- dominant logarithmic distribution of the monthly and annual data, a log-linear link function g(µ) = log(µ) was imple- mented. The smoothing functions of the models were built using regression splines fitted by penalized maximum like- lihood to avoid over-fitting (Wood, 2001). An extra penalty was added to each smoothing term so that each could poten- tially be set to zero during the fitting process, especially in the case of multi-collinearity or multi-concurvity. The inter- action of geographical coordinates was then added to each model (as a two-dimensional spline on latitude and longi- tude) to incorporate spatial trends of the target variable at the regional scale.
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A Study of Risk Analysis Methodology Selection for the External Hazards

A Study of Risk Analysis Methodology Selection for the External Hazards

! One example of deterministic evaluation and evaluation based on engineering judgment is FIVE [9] technique applied to internal fire. For NPPs in Japan, the Comprehensive Assessment of the Safety [10] (so-called ‘Stress Test’) was performed following the accident at Fukushima Daiichi Nuclear Power Plant to evaluate the effects of earthquakes, tsunamis and combined events on the plant safety. The assessment to identify dominant core damage sequences using the Stress Test results is also one example of deterministic evaluations and/or evaluations based on engineering judgment.
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Estimated association between dwelling soil contamination and internal radiation contamination levels after the 2011 Fukushima Daiichi nuclear accident in Japan

Estimated association between dwelling soil contamination and internal radiation contamination levels after the 2011 Fukushima Daiichi nuclear accident in Japan

Since July 2011, the free WBC screenings in Minamisoma city have targeted all Minamisoma resi- dents. 19 Those who wish may undergo screenings once a year. Information on screenings is distributed by the local government and publicised in magazines. These screenings take place at two hospitals in the city that host permanent WBC systems. The WBC machine in Minamisoma Municipal General Hospital is a stereo- scopic machine with two 3×5×16 inch NaI scintillation detectors (Fastscan Model 2251; Canberra, Inc, Meriden, Connecticut, USA). A 30 cm stool was used for residents below 130 cm in height. The WBC device in Watanabe Hospital is in the shape of a chair and has two 3×5×16 inch NaI scintillation detectors (WBC-R43-22458; Hitachi Aloka Medical, Ltd, Mitaka, Tokyo, Japan). The detection limits of both machines are 220 Bq/body for Cs-134 and 250 Bq/body for Cs-137 following a 2 min scan. Calibration was performed on the basis of recom- mendations from the respective companies.
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NUCLEAR ENERGY: THE DEPENDABLE OPTION

NUCLEAR ENERGY: THE DEPENDABLE OPTION

• Handling practice at Yucca Mountain Yucca Mountain is located near Las Vegas in the desert of Nevada. It is 6 mile long 1,200 feet high and it is able to store more than 70,000 tons of radioactive material packed in container. In this waste handling of nuclear waste is carried out by storing spent pallets in carbon or stain steel- zirconium iron containers. Inside yucca mountain 3 tunnels are constructed which are parallel to main rail cart transportation lane which transports spent nuclear pallets packed in container to suitable tunnel depending upon type of waste low, medium and high level waste. And after storage has done entrance of Yucca Mountain in closed with the help of thick anti-radioactive door during this all
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An assessment of radiation doses at an educational institution 57.8 km away from the Fukushima Daiichi nuclear power plant 1 month after the nuclear accident

An assessment of radiation doses at an educational institution 57.8 km away from the Fukushima Daiichi nuclear power plant 1 month after the nuclear accident

(IAEA) has reported lower radiation indoors than outdoors, although indoor radiation rates depend on building mate- rials [18]. The penetrating power of radiation is different depending on the radiation type. b-Rays can be blocked by thin metal such as aluminum, and c-rays can be blocked by lead, iron, and concrete [19]. So, an interior surrounded by concrete with a high shielding effect would have a low dose of radiation. Indoors, however, entrances with people frequently coming and going or areas near windows show high doses of radiation, which suggests that care must be taken not to spend a long time in these locations. It is important to stay inside as much as possible unless you have something to do outdoors, in order to reduce radiation exposure. Moreover, because there was no significant dif- ference in radiation dose rates and count rates depending on floor, we believe there are few differences in radiation effect by floor when in a building surrounded by concrete. Outdoors, there were differences in radiation dose rates and count rates depending on height from the surface, with most measurements at 1 cm above the surface significantly higher than those at 100 cm above the surface. These results match those reported by Ogata, with a tendency for radiation doses near the ground surface to be higher than those in the air [20]. One of the factors for high radiation doses near the surface could be accumulation of radioactive material deposited on the ground. Considering I-131, Cs-134, and Cs-137, with half-life of 8.05 days, 2.06 years, and 30.1 years [21, 22], respectively, cesium will have a larger health impact in the future. Svendsen et al. reported an increase in respiratory diseases after a nuclear power plant accident because dust with high cesium concentra- tions from the soil can often be inhaled [23]. Wearing a mask is considered to be effective to prevent dust inhala- tion, and one way to prevent ingestion of radioactive materials is to avoid outdoor
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Post disaster Policy Decision Making and the Prospects of Human Rights     The Case of Fukushima Daiichi Nuclear Power Plant Accident

Post disaster Policy Decision Making and the Prospects of Human Rights The Case of Fukushima Daiichi Nuclear Power Plant Accident

There were many national concerns in which public participation was critical; however the authorities did not seem to pay sufficient attention. There was a fear of public show down at some point if the government persisted in its attitude to ignoring people’s voices. Particularly, when the government announced plans to revitalize economy including restarting a few commercial reactors, it did not go well with the public. Public sentiments were displayed in antinuclear protests in Tokyo. One particular anti-nuclear protest that took place in Tokyo in July 2012 was reported to have mobilized around 170,000 people according to the unofficial reports and 75,000 people according to the official reports. It occurred as a reaction to government’s indication to restart one or two nuclear power plants for economic reasons to spur growth and save money on import of fuel [32].
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Hemispheric dispersion of radioactive plume laced with fission nuclides from the Fukushima nuclear event

Hemispheric dispersion of radioactive plume laced with fission nuclides from the Fukushima nuclear event

accident on March 11, fission nuclides 131 I (hereafter representing particulate 131 I unless specifically indicating gaseous fraction or all phases. 137 Cs, and 134 Cs in particulate form were measured continuously till early May through aerosol filter samples collected over a sampling network, which includes two offshore-island ground-level stations (Pengchiayu in the East China Sea, and Dongsha in the South China Sea, denoted as PCY and DS, respectively) (25°37′12″N, 122°4′12″E; 20°41′54″N, 116°43′43″E) and two Taiwan inland stations (Nankang at ground level in northern Taiwan and Mt. Lulin at 2862 m high in central Taiwan, denoted as NK and MLL, respectively) (25°2′26″N, 121°36 ′ 50 ″ E; 23°28 ′ 07 ″ N, 120°52 ′ 25 ″ E) (Figure 1). During the sampling period and over the sampling area, the layer of 1.5 km high is under the influencing regime of the pre- vailing northeasterly monsoon. In contrast, the westerlies were the prevailing winds at the elevation of MLL. Sampling
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Obstetric Outcomes in Women in Fukushima Prefecture during and after the Great East Japan Earthquake and Fukushima Nuclear Power Plant Accident: The Fukushima  Health Management Survey

Obstetric Outcomes in Women in Fukushima Prefecture during and after the Great East Japan Earthquake and Fukushima Nuclear Power Plant Accident: The Fukushima Health Management Survey

Objective: The Great East Japan Earthquake (magnitude, 9.0) followed by a large- scale tsunami caused a severe nuclear accident at the Fukushima Daiichi Nuclear Power Plant (Tokyo Electric Company). This study aimed to evaluate the obstetric outcomes in women in Fukushima prefecture during and after the Great East Japan Earthquake and Fukushima nuclear power plant accident. Methods: We collected information for 12,300 pregnant women who conceived during the 9 months before and after the disaster in Fukushima prefecture. The data of the subjects were ana- lyzed according to the conception date for each pregnancy. Results: Among the women who conceived within 9 months before the disaster, adverse obstetric out- comes were not observed. In contrast, in the case of women who conceived within 6 months after the disaster, an increase in the incidence of preterm birth (less than 37 weeks) and low birth weight (less than 1500 g and less than 2500 g) was observed. Moreover, these women showed an increased incidence of medical complications, such as respiratory diseases and mental disorders. Conclusion: The results of the present study show that the occurrence of adverse obstetric outcomes was higher in the women who conceived within 6 months after the disaster than in those who were pregnant at the time of the disaster. The results may be related to emotional stress such as anxiety about the disaster and emphasize the need for continued investiga- tions and careful management of pregnant woman in disaster areas in the future. How to cite this paper: Hayashi, M., Fu-
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A Determination   of Severe Accident Environmental Conditions utilizing Accident Management   Strategy for Equipment Survivability Assessment (P325)

A Determination of Severe Accident Environmental Conditions utilizing Accident Management Strategy for Equipment Survivability Assessment (P325)

Severe accident management encompasses the actions which could be considered in recovering from a severe accident and preventing or mitigating the release of fission products to the environment. Those actions would could be taken were initially clarified by EPRI and were designated as Candidate High Level Actions (CHLAs). In general, Accident Management Program (AMP) for individual plant are developed by considering the spectrum of the CHLAs for specific plant type as well as the anticipated effects associated with the implementation of the high level actions at various stages of an accident. The effects that each action would create are, to varying degrees, dependent upon the extent of damage to the core, RCS, and the containment. Under this consideration, AMGs for KSNPs has being developed and basically seven accident management strategies are developed [7]; Inject into the steam generator, Depressurize the RCS, Inject into the RCS, Inject into the containment, Control the fission product releases into the environment, Control the containment pressure and temperature, and Control hydrogen concentration in containment. For the equipment and instrumentation needed for accident prevention or mitigation, the timing and the condition that these actions would be taken are very important with respect to their functionabilities. These characteristics can be only determined by strategy performance control logic being established for individual plant.
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