Risk Oriented Accident Analysis Methodology Research Articles

Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework

Abstract Nordic boiling water reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core–melt coolant interactions, containment integrity can be threatened by: (i) formation of noncoolable debris bed or (ii) energetic steam explosion. Melt is released from the vessel affect ex-vessel phenomena and is recognized as the major source of uncertainty. The risk-oriented accident analysis methodology (ROAAM+) is used for quantification of the risk of containment failure in Nordic BWR where melt ejection mode surrogate model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine the respective loads on the containment. Melt ejection SM is based on the system analysis code methods for estimation of leakages and consequences of releases (computer code) (MELCOR). Modeling of vessel failure and melt release from the vessel in MELCOR is based on parametric models, allowing a user to select different assumptions that effectively control lower head (LH) behavior and melt release. The work addresses the effect of epistemic uncertain parameters and modeling assumptions in MEM SM on the containment loads due to ex-vessel steam explosion in Nordic BWR. Sensitivity and uncertainty analysis performed to identify the most influential parameters and uncertainty in the risk of containment failure due to ex-vessel steam explosion. The results of the analysis provide valuable insights regarding the effect of MELCOR models, modeling parameters, and sensitivity coefficients on melt release conditions and predictions of ex-vessel steam explosion loads on the containment structures.

IVR-ERVC study of 1700 MW class PWR based on MAAP simulation and coupled analysis

As an effective severe accident mitigation strategy, in-vessel retention – external reactor vessel cooling (IVR-ERVC) has been applied to some advanced nuclear reactors. In the assessment of AP600-like reactor, the risk oriented accident analysis methodology (ROAAM) has been adopted based on a lumped parameter model, including sensitivity studies about key parameters and bounding scenarios during accident states. In the ROAAM method, the quantity of steel in the metallic layer and zirconium oxidized, and decay power is not deterministic. These three parameters result from reasonable analysis as presented in DOE/ID-10460 that has provided reasonable bounding distributions for AP600. This study has performed a MAAP4 simulation for 1700 MW class PWR and focused on core melting progress and bounding distributions of main parameters. The effectiveness of the IVR-ERVC for 1700 MW class PWR has been assessed in a coupled analysis based on a modified lumped model using MAAP simulation results. This analysis has considered the coupled relations about heat transfer between in- and ex-vessel. Also, this study has discussed the applicability of natural convection heat transfer correlations in molten pool and compared the results between the lumped parameter model and the modified lumped parameter model. The sensitivity analysis about the three parameters discussed above has been conducted. The methodology demonstrated in this paper can be used for severe accident analysis and advanced reactor design. The results obtained so far indicated that it is feasible to apply IVR-ERVC strategy to a large scale PWR.

In-vessel retention coolability evaluation for Chinese improved 1000 MWe PWR

After the Fukushima Daiichi accident, the Chinese National Nuclear Safety Administration requires enhancements of safety and mitigation capability under severe accidents for nuclear power plants (NPPs). In-vessel retention (IVR) strategy is proposed to mitigate the consequences of severe accidents for pressurized water reactors (PWR) to satisfy the requirements. Therefore, a proposal with passive cavity injection system (PCIS) is designed for Chinese improved 1000MWe PWR for implementation of IVR strategy. Besides, to condense the steam generated in the cavity head, a passive heat exchange system (PHES) is also designed for passive containment cooling. To evaluate the coolability of the PCIS, a computer code is developed to assess the effectiveness of IVR strategy based on the integrated risk oriented accident analysis methodology (ROAAM) approach. In order to obtain input parameters for assessment, seven severe accident sequences are screened based on Level 1 PRA results and analyzed with integral accident analysis computer code. With the characteristics of the molten pool and key parameters gained from accident analysis, the IVR assessment program is developed with MATLAB code considering parameters of decay power, zirconium oxide fraction, the mass of stainless steel (SS). The influence of thermophysical parameters, critical heat flux (CHF) correlation and mass of metal layer on IVR effectiveness are discussed as sensitivity study for the assessment. Thermophysical parameters have the major effect. Combining the sensitivity study, assessment results show that the IVR strategy for Chinese improved 1000MWe PWR has an acceptable high success probability. However, for engineering practice, detail system design and analysis should be evaluated in the near future.

The development and demonstration of integrated models for the evaluation of severe accident management strategies—SAMEM

This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the generic modelling framework was refined to enable a number of applications. Some recommendations have also been made regarding the applicability of these approaches to existing operating reactors and future reactors. The need for further research and development in the area of human reliability quantification was identified.

Lower head integrity under steam explosion loads

Lower head integrity under steam explosion loads in an AP600-like reactor design is considered. The assessment is the second part of an evaluation of the in-vessel retention idea as a severe accident management concept, the first part (DOE/ID-10460) dealing with thermal loads. The assessment is conducted in terms of the risk oriented accident analysis methodology (ROAAM), and includes the comprehensive evaluation of all relevant severe accident scenarios, melt conditions and timing of release from the core region, fully three-dimensional mixing and explosion wave dynamics, and lower head fragility under local, dynamic loading. All of these factors are brought together in a ROAAM probabilistic framework to evaluate failure likelihood. The conclusion is that failure is ‘physically unreasonable’.

Application of the risk oriented accident analysis methodology (ROAAM) to severe accident management in the AP600 advanced light water reactor

An important part of the AP600 design, as well as of the design certification review by the US Nuclear Regulatory Commission, is devoted to ensuring defense in depth through deep consideration and management of severe accidents. Going beyond the traditional Level 2 PRA, in this article we show how this defense in depth was achieved and demonstrated in a consistent manner between prevention and mitigation, through application of the Integrated ROAAM approach. This requires the up-front integration of probabilistic and deterministic thought, which leads naturally to clear and coherent safety goals as an overall guide toward closure.

In-vessel coolability and retention of a core melt

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the risk oriented accident analysis methodology (ROAAM) and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow for the delineation of the failure boundaries. The technical treatment in this assessment includes: (a) new data on energy flow from either volumetrically heated pools or non-heated layers on top, boiling and critical heat flux in inverted, curved geometries, emissivity of molten (superheated) samples of steel, and chemical reactivity proof tests; (b) a simple but accurate mathematical formulation that allows prediction of thermal loads by means of convenient hand calculations; (c) a detailed model programmed on the computer to sample input parameters over the uncertainty ranges, and to produce probability distributions of thermal loads and margins for departure from nucleate boiling at each angular position on the lower head; and (d) detailed structural evaluations that demonstrate that departure from nucleate boiling is a necessary and sufficient criterion for failure. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is ‘physically unreasonable’ Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings.

On the proper formulation of safety goals and assessment of safety margins for rare and high-consequence hazards

The issue of ‘uncertainty’ is addressed in the special context of assessing and managing risks from rare, high-consequence hazards. It is suggested that, rather than the usual ‘formal treatments’ on how to combine expert opinions that diverge widely, such ‘uncertainty’ must be approached in each case as a research question that encompasses frame of assessment, approach methodology, risk management, and safety goals, with the aim of obtaining resolution in a clear, consistent, and complete manner. This, together with some basic considerations on ‘defense-in-depth,’ and certain practical aspects of communications and synergism needed for resolution (of such uncertainties), leads us to the Risk Oriented Accident Analysis Methodology (ROAAM). The purpose of this paper is to explain these views, to follow them through to the definition of the methodology and its implementation, and to indicate some of the insights gained through the several practical applications available so far.

The probability of containment failure by direct containment heating in Zion

This paper is the first step in the resolution of the direct containment heating (DCH) issue for the Zion nuclear power plant using the risk oriented accident analysis methodology (ROAAM). This paper includes the definition of a probabilistic framework that decomposes the DCH problem into three probability density functions that reflect the most uncertain initial conditions (UO 2 mass, zirconium oxidation fraction, and steel mass). Uncertainties in the initial conditions are significant, but our quantification approach is based on establishing reasonable bounds that are not unnecessarily conservative. To this end, we also make use of the ROAAM ideas of enveloping scenarios and ‘splintering’. Two causal relations (CRs) are used in this framework: CR1 is a model that calculates the peak pressure in the containment as a function of the initial conditions, and CR2 is a model that returns the frequency of containment failure as a function of pressure within the containment. Uncertainty in CR1 is accounted for by the use of two independently developed phenomenological models, the convection-limited containment heating model and the two-cell equilibrium model, and by probabilistically distributing the key parameter in both, which is the ratio of the melt entrainment time to the system blowdown time constant. The two phenomenological models have been compared with an extensive database including recent integral simulations at two different physical scales (1:10-scale in the Surtsey facility at Sandia National Laboratories and 1:40-scale in the COREXIT facility at Argonne National Laboratory). The loads predicted by these models were significantly lower than those from previous parametric calculations. The containment load distributions do not intersect the containment strength (fragility) curve in any significant way, resulting in containment failure probabilities less than 10 −3 for all scenarios considered. Sensitivity analyses did not show any areas of large sensitivity. The feasibility of extrapolating containment loads distributions to most other pressurized water reactors is explored.