Core damage

What If—A Stucly of Severe Core Damage Events,NP 2001,EPRI,PAo Mto,(NM., 1989. 

A significant development of the study was the use of event trees to link the system fault trees to (lie accident initiators and the core damage states as described in Chapter 3. This was a response to the ditficulties encountered in performing the in-plant analysis by fault trees alone. Nathan Villalva and Winston Little proposed the application of decision trees, which was recognized by Saul Levine a.s providing the structure needed to link accident sequences to equipment failure. 

Although the consequences of the high-risk accident sequences may vary from one PSA to another, all PSAs attempt to evaluate realistically, the consequences of hypothetical accident sequences. Expending on the scope of the PSA, these evaluations may include an estimation of the number of latent cancers, the number of immediate fatalities, the probability of core damage, or a number of other consequence measures. 

Table .4.3 3 Estimated Frequencies - Severe Core Damage ...

Fig. 6.3-3 Core Damage Master Logic Diagrams for Indian Point 2 and 3 Nuclear Power Stations.

Core damage can result most likely from heat imbalance. Figure 6.3-3 is an example from the Indian Point PRA that uses heat imbalance to approach completeness. This diagram shows that cote damage may result from either a loss of cooling or excess power (or both). The direct causes of insufficient heat removal may be loss of flow, makeup water, steam flow, or heat extraction by the turbine. Indirect causes are reactor trip or steam line break inside or outside of containment. Cau.ses of excess power production are rod withdrawal, boron removal, and cold water injection. 

The amounts of material released from a damaged plant are usually expressed in fractions of the isotopic quantities in the core. These source terms (meaning source for the ex plant transport) depend on accident physics, amount of core damage, time at elevated temperatures, retention mechanisms, and plate-out deposition of material as it transports from the damaged core to release from containment. This section gives an outline of early source term assessments, computer codes used in calculations, and some comparisons of result.s. 

A long evolving use of PSA was for Anticipated Transients without Scram (ATWS) which extended over 15 years to culminate in NUREG-0460 which was upset by the Salem failure-to-scram incident and the subsequent SECY Letter 83-28. Other special studies have been (a) value-impact analysis (VIA.) studies of alternative containment concepts (e.g., vented containment, NUREG/CR-0165), (b) auxiliary feedwater studies, (c) analysis of DC power requirements, (d) station blackout (NUREG/CR-3220), and (e) precursors to potential core-damage accident.s (NUREG/CR-2497), to name a few of the NRC sponsored studies. 

Precursors to Potential Severe Core-Damage Accidents... 

This study (NUREG/CR-2497) applies PSA techniques using operating experience to identify the high-risk features of plant design and operation. The operating experience base is derived from die licensee event repoits (LERs) to find multiple events that, when coupled with postulated events, lead to plant conditions that could eventually result in severe core damage. 

Potential accident scenarios and flood locations were identified from plant drawings and tlic RHR system fault tree that identifies the equipment and support needed for RHR system operation. The equipment location was correlated with flood areas with consideration for plant features which may impede or divert the flow. The flood scenarios identify the effect on systems required to prevent core damage. Quantification accounts for the rate of rise of the flood relative to the critical level in each specific plant area. The time available for any recovery action is calculated from tiic volume and the flow rate. 

Also, presented are the level-1 uncertainty analysis, results. The MLO mean core damage frequency from internal events is about an order of magnitude lower than that of full power operation. The mean core damage frequency due... 

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. 

Pressure-tubes allow the separate, low-pressure, heavy-water moderator to act as a backup hesit sink even if there is no water in the fuel channels. Should this fail, the calandria shell ilsdf can contain the debris, with the decay heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the concrete containment wall will leak and reduce the possibility of catastrophic structural failure (Snell, 1990). 

The mean core damage frequency from all internal events is 1.8E-4/yr, with an error factor (95% percentile divided by the median) of 5.0. The percentage contributions to the core dai e frequencies are Large Reactivity Insertion 28% Large LOCA, 28% Reactivity Insertion L P ramp, 12% Spurious/normal shutdown, 11% Loss of commercial power, 10% and others, 5 ... 

Fault trees were developed using the IRRAS 2.0 code (Russell, 1988) which allows definition of individual sequences in an event tree, and generation of their cutsets, but does not generate cutsets for total core damage frequency. An in-house code was developed to combine the cutsets uf arictus sequences. Because IRRAS 2.0 was preliminary, use was also made of the SETS code (Worrel. 

At 40 MW operation, the core damage frequency is 3.7E-04/y. The proportion of accident classes is LOCA, 50% beam tube rupture, 27% ATWS, 17% LOOP, 4% and other transients, 2 7. Three minutes of forced flow are not required and large LOCAs with break size smaller than 2.8 inches can be mitigated. 

The point estimate core damage frequencies for the K-Reactor for both internal and external initiators are given in Table 11.3-6. 

The mean frequencies of events damaging more than 5% of the reactor core per year were found to be Internal Events 6.7E-5, Fire 1.7E-5, Seismic 1.7E-4, and total 2,5E-4. Thus, within the range of U. S. commercial light water reactors The core damage frequency itself, is only part of the story because many N-Reactor accident sequences damage only a small fraction of the core. The... 

The confinement maintains its integrity in more than 85% of internally initiated core damage accidents. About half of these failures are from mispositioning of a single switch to fail both the ECCS and confinement isolation. 

The gaseous effluent filtration system was found to operate successfully in 95% of the internally initiated core damage accidents in which the confinement maintains its integrity. 

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