Severe Accident Sequences Research Articles

Aspects of the application of Level 2 probabilistic safety assessment / Aspekte der Anwendung probabilistischer Sicherheitsanalysen der Stufe 2

Level 2 probabilistic safety assessments start from those plant states in which severe core damage is inevitably imminent. The objective is to assess severe accident sequences in order to determine radiological source terms with regard to the environment and their associated frequencies. In most countries using nuclear power, Level 2 PSAs are being considered as an important investigative tool to apply state-of-the-art knowledge on severe accident phenomena in a systematic way. Based on the results of Level 2 PSA, plant vulnerabilities have been identified and plant modifications as well as accident management measures have been proposed and implemented for further reduction and control of the consequences of severe accidents. A research project concerning a generic safety assessment of severe accidents in boiling water reactors was started in Germany in 1993. The study covers full-power states as well as shut-down states. The focus is on the development of a method for the analysis of severe accidents with an integrated probabilistic assessment. It also includes the assessment of accident management measures as well as uncertainty analyses

Assessment of two BWR accident management strategies

Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success.

Modeling of Heat Transfer to Nuclear Steam Supply System Heat Sinks and Application to Severe Accident Sequences

A fast-running computational model has been developed that deals with the nuclear steam supply system heat sink as a two-dimensional slice of steel with its inner and outer surfaces subjected to different thermal and material boundary conditions imposed by such surrounding media as core material, steel layer, water, and gas. This model is generally applicable to two- or one-dimensional heat sinks in the process of heatup and cooldown including liquefaction and resolidification. The numerical model and its solution technique were validated against a set of well-defined initial and boundary value problems. The computer model was applied to analyzing the temperature response of the lower head in a pressurized water reactor large-break loss of coolant accident (LOCA) with ex-vessel cooling. It was of importance to properly account for radiative heat transfer between the two exposed surfaces of the heat sink and the debris bed in the lower plenum, incorporating the physically based view factors, and to allow the heat sink to melt and relocate to the lower plenum. The model was also applied to analyzing the thermal behavior of the lower head in a boiling water reactor large-break LOCA without ex-vessel cooling. It was indicated that the vessel lower head couldmore » undergo a noticeable ablation due to the decay power generated from the debris bed in the absence of external cooling. The computer model was demonstrated to produce consistent results for the applications of practical interest in the severe accident arena.« less

ALWR utility requirements document containment performance requirements

U.S. utilities, with substantial support from international utilities, are leading the industry-wide Advanced Light Water Reactor (ALWR) Program. This program is establishing a technical foundation for the next generation of LWRs through development of a comprehensive set of design requirements for the ALWR in the form of a Utility Requirements Document (URD). The approach in the URD for severe accidents involves two main efforts: (1) accident prevention through intrinsic design characteristics, backed up by reliable safety systems to prevent core damage, and (2) incorporation of design features to ensure severe accident mitigation and containment for the full spectrum of postulated accidents, including core melt accidents. For containment performance, twenty-three severe accident containment challenges are identified, a matrix of design features and operating characteristics is specified to address the challenges, and a preliminary evaluation of the URD indicates that requirements are adequate for addressing each of the challenges. Further, the URD requires evaluation of the containment response to severe accidents. The key conclusions from this effort are: that severe accident challenges are being systematically and explicitly addressed in the design of ALWRs; that margin exists between the loads predicted to result from severe accidents and the Service Level C limits; and that, for a realistic new design basis source term that is expected to bound that from any credible severe accident sequence, the site boundary dose is less than 0.5 rem given the predicted intact containment performance.

蒸気発生器伝熱管破損事象時のアクシデント・マネージメントに係る解析

As a part of the preparation of the database of calculated results on severe accident sequences for emergency preparedness at nuclear power plants, steam generator tube rupture initiated severe accident sequences at a reference two-loop PWR plant have been analyzed using the THALES/ART computer code. Estimation were made of the times of key events such as core uncovery, core melt and reactor vessel melt-through, and an amount of fission products released to the environment. The effects of recovery action on core integrity and the amount of environmental fission products released were studied with regard to the discussion on mitigation of or recovery from the accident by the feed and/or bleed operation in the primary and/or the secondary system. The analysis showed that (1) the ECCS has adequate capability of maintaining core integrity, and (2) the bleed operation in the primary system is effective to suppress the environmental fission products released even if the ECCS did not function.

Simplified methodology for assessing the risk impact of potential performance improvements for the dry pressurized water reactor containment

This paper presents a simplified methodology for calculating the risk benefit associated with potential performance improvements for the dry pressurized water reactor (PWR) containment. The analysis is based on the June 1989 draft NUREG-1150 results for the Zion commercial nuclear reactor. Simplified containment event trees and the large accident progression event trees from draft NUREG-1150 are used to evaluate the effects of potential improvements on the response of the Zion containment to dominant severe accident sequences. Source terms are generated parametrically using the ZISOR code and offsite consequences are calculated with the MELCOR Accident Consequence Code System (MACCS). These results give point estimates of the risk reduction associated with each containment improvement identified by Brookhaven National Laboratory in their draft Issues Characterization Report.

Severe accident analysis of a 1300 MWe PWR to address the hydrogen issue

The behaviour of the potentially large quantity of hydrogen generated during a severe accident has been recognised as an issue of importance since the accident at Three Mile Island. In this article, we describe a severe accident analysis for the Neckarwestheim 2 1300 MWe PWR “Konvoi” plant, performed primarily to investigate the behaviour of hydrogen in the containment, and draw conclusions regarding the need for hydrogen control systems (igniters). The Modular Accident Analysis Program (MAAP) developed by IDCOR in the United States, and the Westinghouse COMPACT multi-compartment containment code were used. The study investigated the generation, release to containment, distribution within containment and potential combustion of hydrogen produced during two severe accident sequences. Results are summarized which show that hydrogen mixing in containment is generally good and that even without hydrogen control systems, hydrogen combustion, although possible, does not threaten containment integrity.

A comparison of the RAFT and VICTORIA computer codes for simple geometries in chemically reactive systems

Two computer codes that predict aerosol formation and transport in the primary circuit of Light Water Reactors are compared for simple problems, to try to establish the main differences between the physical models used in both codes and their effect in predicting severe accident sequences. Besides differences in the nucleation model and in the treatment of chemistry, that are acknowledged by the authors of the codes, important differences in the way the codes treat turbulence were highlighted by this comparison. It is also suggested that the thermochemical databases used for tellurium in the two codes are inconsistent, leading to different predictions of tellurium deposition in the primary circuit.

Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750°F (1783 K) and an oxide debris melting temperature of 4350°F (2672 K). COINTAIN analyses were performed for the debris/concrete interaction occurring without consideration of the possible existence of an overlying pool of water. Results indicate failure of the drywell head seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed.

Molten Material Behavior in the Three Mile Island Unit 2 Accident

The behavior of melts in severe accident sequences affects the nature (composition and fission product inventory) of the debris released from the vessel upon lower head failure in unmitigated accid...

Advanced severe accident response models for BWR application

Boiling Water Reactors (BWRs) and their primary containments have unique features that must be modeled for best estimate severe accident analysis. The BWRSAT Program at ORNL has developed and incorporated into its calculational code suite several advanced response models for application in BWR severe accident analyses. Major features of these models include representation of (1) heat transfer to all in-core structures, (2) the effect of SRV actuations, (3) structural/steam chemistry effects (4) progressive core structural relocation (including “candling”), (5) failure of coreplate and formation of debris bed in reactor bottom head, (6) vessel penetration faiure, and (7) the time dependent egress of core debris from the vessel. This paper discusses these advanced models and presents the results of calculations utilizing the models for the Station Blackout Severe Accident Sequence at Browns Ferry.

Analysis of containment venting at the peach bottom atomic power station

An analysis of the extent to which containment venting would be effective in preventing or mitigating the consequences of severe accidents has been completed for the Peach Bottom Atomic Power Station, Units 2 and 3 (BWR-4s with Mark I containments). The analysis indicates that the effectiveness of venting in preventing containment overpressurization highly depends on the sequence of the severe accident. Containment venting can be effective for several classes of sequences, including transients with failure of long-term decay heat removal and loss-of-coolant accidents with breaks inside the containment. However, based on draft procedures and equipment in place at the time of the evaluation, containment venting has limited potential for further reducing the risk associated with three severe accident sequences currently identified as important risk contributors at Peach Bottom. Means of improving the potential for risk reduction is identified, but their influence on risk is not analyzed.

Human Factors Review for Nuclear Power Plant Severe Accident Sequence Analysis

The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage and preventing radiological release.

A Chemical Equilibrium Estimate of the Aerosols Produced in an Overheated Light Water Reactor Core

The degree of vaporization of light water reactor core materials was estimated using a highly idealized procedure involving (a) specification of the phases that are present for both structural and fuel material, (b) estimation of the vapor pressures exerted by the individual components of each phase, and (c) assuming a degree of vaporization of each phase constituent, allowing equilibration between gaseous and condensed species within the assumed pressure vessel volume. Using this procedure, the aerosol was estimated to consist mainly of silver, indium oxide, cesium hydroxide, and cadmium for pressurized water reactors and cesium hydroxide, cesium iodide, and tellurium for boiling water reactors. If boron is included in the thermodynamic estimate, then boron will significantly alter or dominate the composition of the aerosol in the form of boron oxide and cesium borate. The structural materials make up <9% of the aerosol at 36 to 57 kg, but this figure is in good agreement with estimates from severe accident sequence analysis studies (17 kg) and from Parker (10.7 kg). The SASCHA data are used in NUREG-0772 and give much higher estimates at 295 and 250 kg.

Modes of Heat Removal from a Heat-Generating Debris Bed

In the worst hypothetical accident in a light water reactor, when all protection systems fail, the core could be converted into a deep particulate bed either in-vessel or ex-vessel. The containment of such an accident depends on the coolability of a heat-generating debris bed. Some recent experimental and analytical studies that are concerned with heat removal from such a particulate bed are reviewed. Studies have indicated that bed dryout flux and, therefore, the heat removal rate from the particulate bed increases with the particle diameter (i.e., the permeability) for pool boiling conditions and can exceed the critical heat flux of a flat plate. Bed dryout in a large particle bed (i.e., a few millimetres) was found to be closely related to the ''flooding'' limit of the bed. Dryout under forced flow conditions was found to be affected by both forced and natural convection for mass flow rate smaller than m /SUB cr/ , whereas above this mass flow rate, bed dryout is proportional to the mass flow rate. Recent analyses were found to be in agreement with experimental data; however, additional research is needed to assess factors not accounted for in previous studies (e.g., effect of pressure, multidimensionality, stratification, etc.).more » Based on the expected pressure and particle sizes in a postulated severe accident sequence, a debris bed should be coolable, given a sufficient water supply.« less