T.G. Theofanous

In-vessel coolability and retention of a core melt

Abstract 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.

In-vessel retention of corium at the Loviisa plant

In-vessel retention of corium has been approved to be part of the severe accident management strategy for IVOs Loviisa plant. The approach selected takes advantage of the unique features of the plant such as a low power density, a reactor pressure vessel (RPV) without penetrations at the bottom, and ice-condenser containment which ensures a flooded cavity in all risk significant sequences. The thermal analyses, which are supported by an experimental program, demonstrate that, in Loviisa, the molten corium on the lower head of the RPV is externally coolable with wide margins. This paper summarizes the approach, the thermal analyses and the plant modifications being implemented.

The first results from the ACOPO experiment

The ACOPO experiment simulates natural convection heat transfer from volumetrically heated pools, at a half-scale reactor lower head geometry (hemispherical). Data for Rayleigh numbers of up to 2·1016, from the first round of experiments, are presented in this paper. The results are in substantial agreement with those of the mini-ACOPO proof-of-concept experiment. Moreover, it is shown that these ACOPO results confirm a key component of the in-vessel retention capability for an AP600-like design, as recently established in DOE/ID-10460.

Heat flux distribution from a volumetrically heated pool with high Rayleigh number

Abstract Experimental results are presented on the heat flux distribution at the boundaries of volumetrically heated pools at high enough Rayleigh numbers to be directly relevant to the problem of retention of a molten corium pool inside the lower head of a reactor pressure vessel. The experimental facility, named COPO, is a 2-dimensional “slice”, Joule-heated and geometrically similar in shape (torispherical at 1/2-scale) to the lower head of a VVER-440 reactor. The results show that: the heat flux on the side wall (vertical portion) is essentially uniform; the downward heat flux strongly depends on position along the curved wall; and average fluxes on the side in the downward direction are in agreement with existing correlations, but somewhat underestimated in the upward direction. For the shape considered, the heat flux along the lower curved wall seems to be independent of the presence and extent of the liquid pool (contained by the vertical sidewalls) portion above it.

The coolability limits of a reactor pressure vessel lower head

Configurations II and III of the ULPU experimental facility are described, and results from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Additionally, with Configuration III, we examine the effect of a channel-like geometry created by the reactor vessel thermal insulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related to the observed two-phase flow regimes.

Critical heat flux through curved, downward facing, thick walls

Abstract Experimental data are presented that provide a lower envelope on the critical heat flux distribution over the external surface of a hemisphere submerged in water. The experiment was designed and run so as to provide an essentially full-scale simulation of a reactor pressure vessel lower head. Thus the data, and the correlation derived from them, can be applied directly, and they are supportive of the important, severe accident management idea of retaining the core debris in the reactor vessel by external flooding (“in-vessel melt retention”).

An integrated structure and scaling methodology for severe accident technical issue resolution : Development of methodology

Scaling has been identified as a particularly important element of the Severe Accident Research Program because of its relevance not only to experimentation, but also to analyses based on code calculations or special models. Recognizing the central importance of severe accident scaling issues, the United States Regulatory Commission implemented a Severe Accident Scaling Methodology (SASM) development program involving a lead laboratory contractor and a Technical Program Group to guide the development and to demonstrate its practicality via a challenging application. The Technical Program Group recognized that the Severe Accident Scaling Methodology was an integral part of a larger structure for technical issue resolution and, therefore, found the need to define and document this larger structure, the Integrated Structure for Technical Issue Resolution (ISTIR). The larger part of the efforts have been devoted to the development and demonstration of the Severe Accident Scaling Methodology, which is Component II of the ISTIR. The ISTIR and the SASM have been tested and demonstrated, by their application to a postulated direct containment heating scenario. The ISTIR objectives and process are summarized in this paper, as is its demonstration associated directly with the SASM. The objectives, processes and demonstration for the SASM are also summarized in the paper. The full body of work is referenced.

Natural convection for in-vessel retention at prototypic Rayleigh numbers

The ACOPO experiment simulates natural convection heat transfer from volumetrically heated pools at a half-scale reactor lower head geometry (hemispherical). New data for internal Rayleigh numbers of up to 10 16 are presented, correlated and discussed, in relation to other available correlations. These ACOPO results confirm a key component of the in-vessel retention severe accident management strategy for an AP600-like design, as recently established in DOE/ID-10460.

On the fundamental microinteractions that support the propagation of steam explosions

Abstract This paper makes available the first experimental data on the fragmentation kinetics of hot liquid drops in another liquid (coolant) under the influence of sustained pressure pulses. We observe the effect of “thermal” on “hydrodynamic” fragmentation and micromixing mechanisms, as deduced by the rates and morphology of the resulting particle “cloud”. We show how propagation can be quantified within the framework of a numerical model, and on this basis some interesting interpretations of an experimentally-observed triggered “detonation” in the KROTOS facility (in ISPRA) are offered.

A consistent approach to severe accident management

Abstract The approach adopted for severe accident management (SAM) at the Loviisa nuclear power plant (in Finland) is presented and discussed. The approach includes a number of significant hardware changes and procedures that allow lowering of the lower head thermal insulation and neutron shield assembly, opening of the ice condenser doors, and spraying (externally) of the steel shell of the containment. It is expected that with these changes we can assure in-vessel debris coolability and retention, gradual burning of the hydrogen with good access to the ice condenser, and long term stabilization of the containment pressure, even in the absence of the residual heat removal system. Methodological aspects of demonstrating these SAM objectives, and the status of work in support of related quantifications (of key phenomena), are included in sufficient detail to provide an integrated perspective of the approach taken. The detailed quantifications, separately on each task, will follow, as respective research and quantification programs come to completion.