Damage probabilistic

On August 8, 1985, the U.S. Nuclear Regulatory Commission (NRCf requested the operators of nuclear power plants in the U.S. to perform Individual Plant Examinations (IPE) on their plants. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events (including internal flooding, but excluding internal fire). Generic Letter (GL) 88-20 was issued to implement the IPE request to identify any plant-specific vulnerabilities to severe accidents and report the results to the Commission. ... 

Core damage and containment performance was assessed for accident sequences, component failure, human error, and containment failure modes relative to the design and operational characteristics of the various reactor and containment types. The IPEs were compared to standards for quality probabilistic risk assessment. Methods, data, boundary conditions, and assumptions are considered to understand the differences and similarities observed. 

Partly with the high stakes in mind, changes have been made in U.S. reactor equipment and operation since the TMI accident to reduce the chance of another accident. The results of these changes are reflected in the predictions of probabilistic risk assessments and by a variety of direct performance indicators. For example, in one measure for U.S. reactors, since the pre-TMI days there has been a reduction of more than a factor of 100 in the number of precursors to potential core damage accidents, as reported by the Nuclear Regulatory Commission (Muley, 1990 Belles et al., 1998). 

When damage causation is probabilistic, as in the case of the effects of synthetic substances on human health, torts inappropriately manage risk because it is difficult to link individual exposures with individual health outcomes. 

To apply the probabilistic methods to estimate micro-cracks possible nucleation within a joint metal (fatigue damage processes). 

Harlow, D. G., and Wei, R. P., Probabilistic Aspects of Aging Airframe Materials Damage versus Detection, Proceedings of the Third Pacific Rim International Conference on Advanced Materials and Processes (PRICM 3), M. A. Imam, R. DeNale, S. Hanada, Z. Zhong, and D. N. Lee, eds., Honolulu, Hawaii, July 12-16,1998, The Minerals, Metals Materials Society, Warrendale, PA (1998), 2657-2666. 

Yuen, K.-V, Beck, J. L. and Katafygiotis, L. S. Unified probabilistic approach for model updating and damage detection. Journal of Applied Mechanics (ASME) 73(4) (2006), 555-564. 

Reliability of the reactor system is evaluated on the basis of the probabilistic safety analyses (PSA) by the event tree method on a LOCA, a SGTR, and others. As shown in Table 8, the total occurrence frequency of core damage is approximately 3.5x10 /reactor/year. Contribution of redundancy and diversity of the isolation valves to enhancement of reliability is confirmed by the PSA. Details of the analyses can be found in the reference[15]. TTie total occurrence frequency of the core damage of the MRX is two orders of magnitude lower than that of existing P plants such as Surry, Sequoyah, and Zion. 

The development of these stress criteria for graphite is based on probabilistic considerations and includes a damage model which equates damage (failure) with the inability of the graphite component to accommodate safe shutdown by insertion of control rods or RSC material. Based on the risk contributions from graphite damage to the overall plant risk envelope, reliability allocations for the structural components are obtained. These allocations are then expressed in terms of non-probabilistic structural requirements such as stress-to-strength limits. 

Safety is defined as the absence of factors causing damage to man or the environment. It is a probabilistic value which quantitatively can be expressed in terms of the size of a risk level by an estimation of the probability of different kinds of events and the possible damage from each event (probabilistic analysis of risk). 

Guidelines for the probabilistic safety analysis have been issued recently. In the guidelines, the analysis of initiating events in other operational states than full power is recommended if essential contributions to the total core damage frequency are to be expected. Efforts have therefore concentrated on limited shutdown analysis for a typical PWR and BWR. A low power and shutdown analysis for one PWR has been completed, one for a BWR will be finished soon and a third, major PWR analysis has just been started. 

Probabilistic analyses are applicable to Levels 1, 2 or 3 (IAEA, 1992, 1995,1996), because they examine the events up to core damage, up to the evaluation of radioactivity releases from the plant, or up to the external radiological consequences. 

It is practically impossible to perform a probabilistic treatment of the vessel failure in severe accidents, that is where there is major damage and core melt, because of a lack of sufficient data on the phenomena and on their probabilities. 

An increase of thickness up to 1.8 m guarantees an absence of damage due to the simultaneous explosion of the normal weapons carried by a fighter aircraft (missiles), but not of the possibly carried bombs (which is justified on a probabilistic basis if the bombs are not triggered to explode). This thickness also offers protection against other types of impacts, such as an obhque one due to the separation of an engine and that of a missile due to the explosion of a nuclear plant turbine (for which in general 80 cm are sufficient). 

The probabilistic safety objectives (10 per reactor year for the damage of the core and 10 per reactor year for releases higher than the limit ones, according to the lines of INSAG which seem to represent a common basis of the most recent regulatory positions. 

The SARDB consists ofhundreds ofcalculationresults for significant core damage scenarios by using integrated severe accident analysis codes, for example, MAAP (FAl, 1994) or MELCOR (USNRC, 1990). It includes most of the high frequency accident sequences which are selected from the plant damage state event tree of a level 2 probabilistic safety analysis... 

From those, a criterion closer to a direct evaluation of the probabilistic safety assessment results was derived, which states that the expected frequency of the large early releases from a NPP should be less than lOE-6/year. In addition, the core damage frequency should be proved to be below lE-4/year. 

Technical criteria are generally quantitative (probabilistic) and mostly on lower levels (subsidiary). They typically concern core damage, unacceptable release, and unacceptable health risks. In later years, some countries have defined separate criteria to address robustness in defence in depth, e.g., by having a separate criterion for reactor containment integrity. 

Figure 1. Probabilistic criteria defined for the frequency of reactor core damage accident. 

Because of the variation in assessment factors and their uncertainty (lightning characteristics, attachment locations, damage models, etc.), a probabilistic or statistical approach can be used to express a relationship for the probability of failure of a critical chemical equipment item. For a single equipment item, this relationship can be expressed as ... 

Probabilistic envelope damage of individual apartment models is assessed through Monte Carlo... 

ABSTRACT When a fire Probabilistic Risk Assessment (PRA) is modeled and quantified by using predeveloped internal PRA model, if components are damaged by a fire, the basic event values of the components became True or one (1), which removes the basic events related to the components from the minimal cut sets, and which makes it difficult to calculate accurate component importance measures. Thus, a new method to accurately calculate Fussell-Vesely importance measure in fire PRA is recently introduced. However, the new method has a drawback when the failure probability of the damaged component is small. Thus, another new method could be proposed. Two methods are compared, and the condition in which each method is accurately applicable is derived in this paper. 

Event tree models They describe sequences of events (hke, for instance, the failures of safely system functions) which may lead to a damage state and provide a corresponding probabilistic assessment. 

Event-tree models in a Level 1 PSA generally account for the order of demands of safety system functions at set points and for the (stochastic) failure behaviour of the required functions. It is common practice to consider just two alternative states at each set point, namely required function is successful and required function fails . No satisfactory consideration is given, for instance, to situations where technical safety systems which are successfully started fail to function with the required capacity and / or fail to run within the required mission time. What is the consequence, if sequences accounting for stochastic failure times of safety system functions are not considered Is the resulting spectrum of event sequences still sufficient enough to obtain an adequate probabilistic assessment for (core) damage states How reahstic are the probabilistic assessments for damage states derived from static event tree models ... 

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