Use of instruments in severe accident guidance is focused on cooling of the reactor core and heat removal from the containment. The large uncertainties and extreme conditions involving the severe accident render the measurement and interpretation of the data difficult enough. There presently exist some parameters relevant to monitoring the accident initiation and progression and to preparing necessary measures and strategies to prevent further aggravation and to mitigate the consequence of the accident at hand. One of the accident management guidance parameters for the KSNP (Korean Standard Nuclear Plant) is the core exit temperature (CET). Though the fuel rod temperature is perhaps the most important parameter to determine the phase of accident progression, there is currently no means to read it directly from the core. A best alternative, thus, is to guess the fuel temperature from the measured CET data. The severe accident management guidance (SAMG) suggest that the transition criterion from the emergency operating procedure is the CET of 650 K. The fuel temperature begins to escalate from the time of core uncovery. When the temperature reaches 930 K, hydrogen is generated by oxidation of the cladding whose reaction is accelerated drastically with temperatures exceeding 1500 K. Because the oxidation reaction is exothermal, the core temperature increases rapidly. Although the CET is lower than the fuel temperature, the increase rate will tend to follow that for the fuel rod temperature. In this paper, the focus is placed on the plant damage state during a severe accident. The accident initiator is a small-break loss-of-coolant accident (LOCA). The MAAP4 calculations were carried out to analyze the plant state and accident sequence such as the core uncovery time, the CET, and the primary system pressure. It is concluded that the CET will prove to be a reasonable criterion for recognizing severe core damage. However, when the temperature exceeds 950 K the CET reading may not be reliable so that one needs alternatively to infer the degree of core damage by reading the reactor water level and the hydrogen generation rate.
After Fukushima accident, it becomes very important issue to investigate BDBAs (Beyond Design Basis Accidents) more accurately in NPP (Nuclear Power Plant) design and operation. In such a severe accident condition, the molten core material forms an internally heated debris bed and eventually becomes a melted pool of corium, which will cause or induce thermal and mechanical loads to the reactor vessel wall and then can make penetrations leading to the failure. A good understanding on the mechanical behaviour of RVLH (Reactor Vessel Lower Head) under the condition is essential for verification of the integrity and improvement of accident mitigation strategies.
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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 ; 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.
According the design 8 passive autocatalytic recombiners (PARs) are installed. They are capable, based on fully passive principle, to recombine hydrogen and oxygen into water steam. According to the performed analyses they are sufficient for recombining the hydrogen, which will be generated during in-vessel phase of the severe accident. In 2013/2014 are installed 15 additional PARs per unit, and the origin of these PARs is from 3 and 4 unit of KNPP. With these additional PARs they are sufficient for recombining the hydrogen, generated during outside vessel phase of severe accident. Scheme of PARs are shown in Figure 3.
Structural components like Steam Generator (SG) tubes, Reactor Coolant Pump (RCP) seals, hot leg, Pressuriser surge line and Reactor Pressure Vessel (RPV) are likely to experience high temperature under postulated severe accident conditions. A high pressure postulated transient like Station Blackout has selected for analyzing the mentioned components thermal-hydraulic loading for VVER1000 (V320) Plant. The analysis has been carried out with severe accident code ASTEC which is used extensively for severe accident analysis for French PWRs and other European Nuclear Power Plants. The estimated temperature and pressure distribution in the reactor coolant circuit show that likelihood of failure is very high for Pressuriser surge line due to frequent operation of Safety Valves mounted on the Pressuriser where as failure possibility is found to be less for SG tubes, RCP seals, hot leg and RPV. These estimation are necessary prior to molten core slumping into lower head of the RPV as failure of one of the components will alter the course of the accident. The alteration includes system depressuring followed by less inventory/steam available for hydrogen generation and converting the possibility of high pressure melt ejection into a low pressure melt ejection scenario. The change in scenario will alter the load on the ultimate barrier i,e the Containment.
The results in Figures 1 and 2 are for scale model tests of two concrete containments. Rebar and concrete crack spacing, and aggregate size can affect the leakage rates in full size containments. However, based on the results of testing and analyses presented above, it is reasonable to conclude that all containments start to leak once the rebars and liner plate yield. In addition, leakage becomes excessive once the strains in the reinforced and prestressed concrete containments reach about 2 and 1 percent respectively. Based on information of the containment model test results and analyses data presented in Figures 1 and 2, it is reasonable to assume that containment leakage of 10 percent of containment volume per day when rebars and liner plate yield. Similarly, leakage rate of 100 percent can be conservatively used in severe accident analysis when the containment global strains are 1-2 percent. Figure 3 shows the proposed relationship between leakage rate and internal pressure based on these assumptions. It is clear from Figure 3, that containment opening area (leakage) increases at a on a lognormal scale with pressure. Uncertainty in the leakage rate can be accounted for by using a variation of 50 percent in the rate as shown in Figure 3.
It is difficult to assess rupture behavior of the lower head of reactor pressure vessel in boiling- water-type nuclear power plants due to severe accident like Fukushima Daiichi. One reason is that boiling water reactor (BWR) lower heads have geometrically complicated structure with a lot of penetrations. Another one is that BWR lower head is composed of various types of materials of RPV, weld-overlay cladding, control rod guide tubes, stub tubes, welds, etc. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional thermal-hydraulics (TH) and thermal-elastic- plastic-creep analyses. To conduct such analyses, we have continued to obtain experimental data on creep deformation in high temperature range. In this study, we performed creep damage evaluations based on developing analysis method by using detailed three-dimensional model of RPV lower head with control rod guide tubes, stub tubes and welds. Creep damage evaluation based on four types of damage criteria of “Considere”, strain, Kachanov, and Larson-Miller-parameter (LMP) were made by using experimentally determined parameters. To investigate the effects of the debris depth and heat generation locations on failure behavior of lower head, we conducted parameter studies varying analysis conditions related to relocated molten core. From the analysis results, we discussed the outflow paths of the relocated molten core to the containment, and it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.
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Then, the global 3D model provides the boundary conditions of the local model (the equipment hatch) which gives in its turn the boundary conditions of the restricted model "sleeve and flanges". CPU cost for the severe accident with the local model is about ten time smaller than for the global model. So, many calculations could be conducted, modifying some mechanical parameters, boundary conditions, dimensions of the local models, characteristics of the screw... Numerical results do not vary significantly when dimensions of the local model increase, this point shows the validity of the method.
To obtain key insights into the underlying severe accident phenomena expected in an SFP, plant-specific analyses for two potential accident scenarios have been performed using the MELCOR code. For this purpose, the plant-specific MELCOR inputs were developed for spent fuels, racks, pool, and SFP building. Table 6 summarizes the key accident phenomena obtained through the present analysis. In the case of a loss of cooling accident, the time to the pool uncover is obviously dependent on the level of decay heat load. In the case of a loss of pool inventory, the break size and location could be a dominant factor in determining the times to the key phenomena and further accident progression.
The general objective of the ISP-46 is to assess the capability of the computer codes to model, in an integrated way, the physical processes taking place during a severe accident in a pressurized water reactor, from the initial stages of core degradation through to the behaviour of the released fission products in the containment. The codes are supposed to be used in a similar manner as they would be used for plant studies, employing standard models and options as far as possible, with representations of the facility and similar details as used for plant studies. The recommendations for the appropriate noding of the Phebus facility are given in the ISP-46 specification , where the number of nodes in different parts of the facility is prescribed. A number of codes were used by the ISP-46 participants, simulating the Phebus facility as a whole or part of it . The Joef Stefan Institute Reactor Engineering Division participated in the ISP-46 with two versions of the MELCOR 1.8.5 computer code (QZ and RE) ( and ) (bundle and circuit part) and with the CONTAIN 2.0 computer code  (containment part).
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An effective countermeasure for Severe Accident-Management (SAM), consisting of an additional base- isolated Diesel Generator System was built at Beznau NPP in June, 2013. Aim of this paper is to describe scope, design-basis as well as technical aspects related with the adopted base-isolation system. The decisional process for the choice of the appropriate base-isolation system will be presented. Further, aspects of seismic robustness will be addressed and discussed. Finally, a cost-to-safety effectiveness of this measure with respect of Core Damage Frequency (CDF) will be shown with the aim to provide some decision making basis for similar projects at other NPP sites.
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This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure decomposition event tree (DET) which is one of the containment event tree (CET) top events in a reference plant of this study. An uncertainty analysis of a containment pressure behavior during severe accidents has been performed for the optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to an in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to 1) develop severe accident scenarios for which containment pressure loads should be performed based on Level 2 PSA, 2) identify severe accident phenomena relevant to early containment failure, 3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, 4) prescribe likelihood descriptions of the potential range of these parameters, and 5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions.
This paper discusses the integrity of the calandria vessel of a CANDU reactor under severe accident loads. The potential failure modes of the vessel are identified. They include thermal creep failure, overpressure failure and mechanical overload failure. Thermal creep failure of the vessel can occur if film boiling occurs over a significant area at the bottom of calandria vessel, thereby causing the vessel wall to heat up. Overpressure failure can occur if large, sustained rapid steam generation associated with relocated core debris into the moderator fluid occurs and the generated steam cannot be adequately relieved through relief ducts. Overload failure would occur if the mass of relocated core debris is sufficient to produce stresses in the vessel wall that exceed ultimate tensile stress. The limits associated with these failure modes are evaluated and used to generate representative likelihoods of calandria vessel heat loads..
The consequence analyses performed in this study employ two codes developed at Sandia National Laboratories, MELCOR and MACCS. MELCOR 1.8.6 (Gauntt et al., 2005) is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is under ongoing developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heat-up, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. The MACCS code (Chanin et al., 1990a, 1990b, and 1990c) was developed for the U.S. Nuclear Regulatory Commission (NRC) to evaluate offsite consequences of hypothetical severe accidents at nuclear power plants (NPPs) based on calculated or assumed fission product releases. The code is designed to take input directly from the fission product releases calculated by MELCOR. MACCS also incorporates geographic, demographic, and meteorological data for a given plant, as well as assumptions concerning biological uptake. The output results can be expressed in terms of health consequences. However, the present study only considers the health consequences in terms latent cancer fatality risk.
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In the event of severe flooding beyond design basis, water may enter fuel pool. The consequence analysis for this condition has been carried out which indicates it is a safe situation, as cooling of the pool water is not hampered. For any flooding more than the margin available, water will enter into fuel pool. As electrical sub-station, DG sets, pool water cooling system pumps and heat exchangers are housed at floor level (FFL) and batteries are at 1m from FFL, these systems will be affected. The area lighting also will be affected. Since control room and UPS are at floor-1 (6m above FFL), some of the important parameters like pool water level and temperature will be available for at least 2 hrs from power failure. But no catastrophe will take place since water acts as cooling medium. Only minor contamination (upto 500 µµCi/ml level) will take place in the flooded area
Evaluation of ultimate load capacity (ULC) of containment is carried out to establish pressure resis- tance margin over its design pressure. ULC is calculated by analytical approaches considering realistic be- haviour of materials, unlike the conservative elastic behaviour assumed in design stage. To gain confidence in analytical predictions, analysis methodology and models are validated through experiments on scaled model of containments. The AERB-USNRC Standard problem Exercise - 3 (SPE-3) on ”Performance of Containment Vessel Under Severe Accident Conditions” was framed on the basis of 1:4-Scale pre-stressed concrete containment vessel (PCCV) model tests performed at Sandia National Laboratory, USA. Scope of the exercise included analytical assessment of ultimate load capacity of PCCV and characterization of its leakage behaviour as a function of pressure and temperature.
Heating tests and in-plane tensile tests were conducted to investigate the structural properties of steel plate concrete walls with openings which simulated a steel plate reinforced concrete structured containment vessel (hereafter abbreviated as an “SCCV”) under high temperature and internal pressure in the event of a severe accident. The test specimens were 1/12-scale models of approx. 20m long x 20m wide x 2m thick wall with openings of the SCCV. The wall thickness of the specimen was 167mm, and the steel plate thicknesses of both sides of the wall were 1.7mm. Test parameters were the size and the number of openings. No-opening specimen has no openings. Large-opening specimen has one large circular opening. Group-opening specimen has six small circular openings. To reinforce the steel plate of the circular opening, the steel plate around the opening was thickened to 3.4mm i.e. twice the thickness of the steel plate of a general part. To reproduce the temperature distribution in the event of a severe accident (hereafter abbreviated as an “SA”), one side of the specimen was heated to a fixed temperature of 300°C, while the other side was cooled to a fixed temperature of 74°C. Afterwards, a tensile force equivalent to the internal pressure in the event of an SA was applied to the four sides of the specimen. Test results showed that the load versus mean strain relationship of the large-opening specimen and group-opening specimen have almost the same properties as no-opening test specimen. Besides, no damage was observed at the opening edge. Thus, the tests confirmed that the opening reinforcement method was appropriate. Moreover, simulation analyses of individual tests were conducted using a finite element method, which confirmed that the analytical results closely corresponded to the results of experiments.
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PDSA 4: In the initial analysis period of the project it was identi ﬁ ed that barriers to productive communication in ﬂ uenced the recognition and timely treatment in severe sepsis. These potential communication failures ranged from poor communication due to staff not knowing which nurse/doctor was looking after the patient, to delays in treatment due to staff not communi- cating that treatment had been prescribed. The project team through observations and feedback from all grades of staff found that there were times when the staff did not know the skills or grade of other staff and also which staff were assigned to which patients. This obviously could have impacted on staff and patient safety. To improve the communication between the team it was decided to look at the introduction of a safety huddle at the start of shifts. A number of styles were investigated and trialed using feedback from the staff to guide the project team to formulate a set standard for the depart- ment. The introduction of the ‘ pre brief ’ comprised of a huddle in each area of the department at the start of shifts, allowed nursing and medical teams to be intro- duced to the name, role and relevant skills of each member of the team, such as if nurse was able to do venepuncture or administer intravenous medications. To support the introduction of pre briefs, name boards were introduced in the individual patient rooms which displayed information of the named nurse and consult- ant in charge of the patients care. This ensured the doctor or allied health professional knew the named nurse who they should communicated with in regards to the management plan and treatment prescribed for the patient. Compliance with the number of completed pre briefs and name boards was undertaken, alongside quali- tative data from staff feedback and questionnaires to understand the impact on team communication.
The study has revealed that there is significant bur- den of head injury in the study area. Young males were found to be affected more often. A similar find- ing has been reported in many previous studies (2, 3, 6, 12). This might be attributed to the existing culture that males are mainly engaged in outdoor activities including work and social gathering while the women spend most of their time indoors, and may have limited outdoor activity. Males may travel more and show more aggressive behavior which leads them to be involved in fighting and road traf- fic accidents. The commonest cause for head injury was found to be fall down accident. Previous studies in other countries had shown similar observations (13). Fall down has been identified as the main cause of head injury in children under 15 and above 60 years of age. Other studies also showed that fall down accidents are associated with the extreme age groups (less than 4 and more than 65 years) most of which are domestic accidents (14, 15,16). Fall down accident can also be work related. Though much fo- cus is given to Road Traffic accident, as interper- sonal violence and fall down accidents have also been found to be the causes for head injury attention should be given to them by the government.
The methodology adopted in this study is based on data from various govt. sources namely; Hyderabad Traffic Police, Hyderabad Metropolitan Development Authority (HMDA) reports are the main sources for Accident data. The collected accident data includes the time of occurrence of accident, type of accident, vehicles involved in the accident, location of accident, number of affected persons etc. The collected data will be tabulated and the general analysis has to be carried out. The General analysis includes like total number of fatal and non-fatal accidents and total number of accidents by yearly wise, monthly wise, daily wise etc.