Light Water Reactor Safety: The Development of Advanced Models and Codes for Light Water Reactor

375 Pages·1995·4.92 MB·English

by  J.N. Lillington

Light Water Reactor Safety

The Development of Advanced

Models and Codes for Light Water

Reactor Safety Analysis
Light Water Reactor Safety

The Development of Advanced

Models and Codes for Light Water

Reactor Safety Analysis

J.N. LILLINGTON

AEA Technology
Winfrith Technology Centre
Dorchester, Dorset, U.K.

1995
ELSEVIER

Amsterdam - Lausanne - New York - Oxford - Shannon - Tokyo
ELSEVIER SCIENCE B.V.

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L1brary of Congress Catalog1ng-1n-Pub11cat1on Data

L1111ngton, J. N.

Llght water reactor safety the development of advanced models

and codes for 11ght water reactor safety analysls / J. N.

L1111ngton.

cm.

Inc l udes blb l lograph1ca l

ISBN 0-444-89741-0

1.

references and 1ndex.

(acld-free paper)

Llght water reactors--Safety measures.

TK9203.L45L55

I. Tltle.

1995

621.48'35--dc20

95-2322

CIP

ISBN: 0444 8974 1 0

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v

PREFACE

A large proportion of the nuclear reactors in operation worldwide are Light Water Reactors
(LWRs). There are two principal designs, Pressurised Water Reactors (PWRs) and Boiling
Water Reactors (BWRs). Over the past 10-20 years, much experimental and theoretical
research has been carried out in Europe, the USA and elsewhere to ensure safe operation and
to understand the behaviour of these plants under various accident conditions. This has
included a significant level of severe accident related research, particularly since 1979 when
a core melt-down occurred at the Three Mile Island - Unit 2 PWR in the USA.

It is of paramount importance that the core in a nuclear reactor should remain adequately
cooled at all times. Even after a successful shut down, heat continues to be produced from
the decay heat of unstable isotopes in the fuel. One of the main objectives of reactor safety
research is to consider ways in which core cooling can be achieved and the consequences if
it is not achieved.

Thus, in the case of a LWR, the physics of water/steam flows, associated heat transfer and
other phenomena (e.g. hydrogen production and fuel degradation, in severe accidents) that
could occur under accident conditions need to be adequatel y understood. Many complex and
expensive experiments have been perfonned and sophisticated analysis techniques (large
computer codes etc.) have been developed to provide the necessary infonnation.

The main purpose of this book is to provide a review of and references to the main activities
that have been carried out towards the development of advanced mechanistic models/codes
for this LWR Safety Analysis. The book will describe the state of the art and discuss the likel y
future direction of research in this area.

This objective will be achieved by summarising the basic features of LWRs that impact on
safety, the key accident scenarios and physical phenomena, the major experiments, the
resulting models/codes developed and their validation together with some representative
plant calculations that have been perfonned with the codes.

The book opens with a brief historical review of the key events from the inception of civil
nuclear power soon after the Second World War up to the present time. The prominant safety
studies in support of Light Water Reactor design and licensing, the major areas of technical
research and nuclear safety research objectives are among the main items summarised. These
latter objectives have helped to define modelling requirements which have then driven
research to provide experimental data and tools for analysis.

The principal features of the designs of both types of Western LWRs i.e. PWRs and B WRs,
relevant to the safety issues presented in this book, are given. These include the elements of
the reactor coolant system (PWR), steam cycle (BWR), the core and pressure vessel, the
containment and the main protection systems.

The main emphasis of the book is on the PWR but where appropriate some mention is made
of the BWR. Many of the important modelling requirements are common to both designs.
VI

Preface

Accident conditions are usually classified within various categories. These categories range
from classes of relatively trivial faults expected within the working life of the plant, through
to design basis accidents and the very low probability beyond design basis (severe) accidents.
Within these categories there are various types of accident condition. A range is described
which is intended to span the scope of relevant phenomena.

Integral experiments are performed for two main reasons: (a) to identify important phenomena expected to exist within a certain accident scenario and (b) for code validation. The main
integral programmes addressing both thermal-hydraulic and severe accident issues are
briefly reviewed. Separate�ffects experiments are referenced where appropriate in the
chapters devoted to individual modelling areas.

Thermal-hydraulic model development has been a major activity during the last twenty years.
As understanding and analysis techniques have improved the trend has been to produce "best
estimate" models/codes with some justified estimate of uncertainty. These are replacing the
conservative or pessimistic models/codes which supported safety cases hitherto. Some of the
important models available in present best estimate system codes are described.

The response of a nuclear power plant under many forms of accident condition will depend
on its response to abnormal heat and pressure loads. Models for component (fuel rods,
structures etc.) heat conduction and heat transfer coupling to the thermal-hydraulics are
reviewed. Mechanical response models in response to heat loads are also swnmarised.

Under severe accident conditions there are a large number of additional phenomena(
compared with the phenomena present in design basis accidents) which also require
modelling. Core melting, reactor coolant circuit failure and the threat to containment are
possible events that need to be considered. It is these events that have attracted the most
attention in recent years. A large proportion and the emphasis of the book therefore
concentrates on severe accident modelling.

At sufficiently high temperatures, oxidation of core components, particularly Zircaloy fuel
rod cladding results in the production of significant hydrogen. This is an extremely important
reaction in relation to the progression of many severe accident conditions. Considerable heat
is evolved during the oxidation process resulting in strong positive feedback and further
increase in temperature. Nwnerous materials reactions occur, resulting in loss of geometry
through the production of low melting point eutectics. Melting points of indi vidual constituents present in the plant are much higher. A chapter is devoted to all these various interactions.

Both mechanistic and parametric models have been produced for predicting the meltdown
and degradation of the reactor core under these extreme conditions. The sequence of events
is affected by the formation of the different eutectics particularly those formed between the
metallic Zircaloy and other core components.

During a melt progression sequence there is the potential for molten material to fall into water
giving rise to an energetic interaction. Steam explosion phenomena and research are briefly
summarised for both in-vessel and ex-vessel melt/water interactions.
Preface

vii

If the debris contacts water varying degrees of fragmentation may occur. Whether such debris
is coolable or not is a key safety issue. If the volume to surface area ratio of the debris is too
large this may not be the case. If the debris becomes too finely fragmented the potential for
steam explosions increases.

During an in-vessel melt progression phase, debris may interact with the vessel and the vessel
internals. There is a particular interest in the way instrument penetrations may be attacked in
the lower vessel head of a PWR, providing a mechanism for vessel failure.

The release and transport of fission products in the primary circuit is briefly covered. While
thermal-hydraulics, fluids and thermal transfer are the main concerns of the book, some
discussion of fission product issues both in the primary circuit and containment is included
for completeness.

Material behaviour in the cavity is discussed. The extent to which debris might be swept out
of a cavity in a high pressure release from the vessel determines the threat of direct
containment heating and potential for early failure of the containment.

Core/concrete interactions provide an important source mechanism for fission product
release to the containment. Fission products would be carried along with copious production
of gases e.g. steam, carbon dioxide and aerosols. These gases may also be reduced by metallic
corium resulting in the production of further flammable hydrogen (and carbon monoxide).

Containment thermal-hydraulics and the transport of fission products within the containment
are the subjects reviewed in the final chapters concerned with modelling. New phenomena
associated with the operation of engineered safety features such as ice condensers and sprays
are also addressed.

A chapter is given to thermo-physical models. Densities, thermal conductivities, specific
heats etc. are required for the important materials e.g. Zircaloy, fuel, stainless steel and
concrete structures. These properties are also required for the material compounds e.g.
oxides, eutectics and for water/steam and various gas properties.

Many codes for LWR safety analysis are now being developed. Emphasis here is given to the
system codes developed by the United States Regulatory Commission (USNRC} and its
contractors. These codes are widely used worldwide and provide a state-of-the-art capability
for primary circuit thermal-hydraulics, in-vessel core degradation and primary circuit severe
accident modelling, fission product behaviour and containment thermal-hydraulics.

Some representative simulations of various integral experiments are shown in a later chapter
to give an overall impression of the adequacy of current predictions, compared with
experimental data.

The TMI -2 accident is judged to be a sufficiently important subject to merit a chapter by itself.
This accident provides unique data at full plant scale on the nature of core degradation and
melt phase progression.
Vlll

Preface

Towards the end of the book selected code calculations of certain representative accident
sequences are briefly presented. Plant studies are the ultimate goal of the model development
and the experimental research.

Accident management modelling requirements are briefly discussed. These are providing
new challenges to the system codes at the present time. Boundary conditions may be
complex: there are also difficulties in modelling certain new phenomena e.g. flooding of a
degraded core.

The final chapter is concerned with Advanced LWRs. These put greater emphasis on passi ve
safety systems and new phenomena are present. New experimental programmes are planned
and underway and further systems code development will be influenced by Advanced LWR
safety concerns.