No C 250/6 Official Journal of the European Communities 19.9.83 
Proposal for a Council Decision adopting a research programme on reactor safety 
(1984-1987) 
(Submitted by the Commission to the Council on 17 June 1983) 
THE COUNCIL OF THE EUROPEAN 
COMMUNITIES, 
Having regard to the Treaty establishing the Euro­
pean Atomic Energy Community, and in particular 
Article 7 thereof, 
Having regard to the proposal from the Commission 
submitted after consultation of the Scientific and 
Technical Committee, 
Having regard to the opinion of the European 
Parliament, 
Having regard to the opinion of the Economic and 
Social Committee, 
Whereas the Council adopted on 22 July 1975 a 
resolution on the technological problems of nuclear 
safety; 
Whereas the Council adopted on 18 February 1980 a 
resolution on fast breeder reactors; 
Whereas the Council has, by Decision ..., approved 
the structures and procedures for the management 
and coordination of Community research, develop­
ment and demonstration activities; 
Whereas the Council, by Decision ..., adopted a 
framework programme defining a European scien­
tific and technical strategy; 
Whereas the Commission has communicated to the 
Council an action programme on the development 
of nuclear fission energy; 
Whereas the implementation of research pro­
grammes concerning nuclear safety is one of the 
principal means whereby the Commission can con­
tribute to the non-hazardous production of nuclear 
energy and to the protection of workers, the general 
public and the environment; whereas this imple­
mentation covers the specific options and objectives laid down in the framework programme and in the 
action programme on the development of nuclear 
fission energy; 
Whereas the Council, by Decision ..., adopted a 
direct action research programme on reactor safety; 
Whereas it is advisable to supplement the direct 
action programme on reactor safety with a shared-
cost action programme making use of the skills and 
plant available in the Member States, 
HAS DECIDED AS FOLLOWS: 
Article 1 
A research programme on reactor safety, as des­
cribed in the Annex is hereby adopted for a period 
of four years from 1 January 1984. 
Article 2 
The amount required for performance of this pro­
gramme is estimated at 81300 000 ECU with an 
estimated staff complement of 17 officials. 
Article 3 
The programme defined in the Annex may be 
reviewed at the end of the third year in accordance 
with the appropriate procedures. 
Article 4 
In the implementation of this programme the Com­
mission shall be assisted by the Management and 
Consultative Committee 'Nuclear Fission', the 
terms of reference and composition of which were 
defined in Decision ... (structures and procedures 
for the management and coordination). 
19. 9. 83 Official Journal of the European Communities No C 250/7 
ANNEX 
SHARED-COST RESEARCH PROGRAMME ON REACTOR SAFETY 
The programme consists of theoretical and experimental research, to be carried out in collabora­
tion, concerning accident prevention, the detailed study of accidents and their consequences and 
the methods employed for probabilistic risk assessment. 
The programme consists of two sections: 
(a) one section concerning the safety of light-water reactors in which the following specific 
points will be studied: 
— human factors and man/machine interaction; 
— protection of nuclear installations against gas-cloud explosions; 
— mechanical and materials problems posed by the steel components of light-water 
reactors; 
— thermohydraulics and severe fuel degradation during loss-of-primary-coolant accidents; 
— problems of the distribution, combustion and monitoring of the hydrogen produced 
during and accident; 
— source term caused by fission products in severe accidents; 
— atmospheric dispersion of fission products after an accident; 
— methods used for probabilistic risk assessment; 
(b) one section dealing with sodium-cooled fast reactors in which the following specific points 
will be studied: 
— instrumentation, control and protection; 
— transient analysis; 
— integrity of components and structures; 
— safety aspects of sodium technology; 
— fuel behaviour in transient conditions and post-failure phenomena (in-pile experi­
ments); 
— fission product transport in severe accidents; 
— movement and interaction of molten materials in severe accidents. 
The programme will be implemented under contract. 
No C 250/8 Official Journal of the European Communities 19.9.83 
PROPOSAL FOR A SHARED-COST RESEARCH PROGRAMME 
ON REACTOR SAFETY (1984-1987) 
(Technical appendix) 
FIRST PART 
1. PROPOSAL FOR A SHARED-COST RESEARCH PROGRAMME 
ON LIGHT-WATER REACTOR SAFETY (1984-1987) 
INTRODUCTION 
Light-water nuclear power stations attained industrial and commercial maturity in the early 
1970s. At present, over 200 power stations of that type are in operation throughout the world and 
over 130 are under construction. In the Member States of the Community, about 40 power sta­
tions are in operation and 30 under construction, which represents an installed nuclear power 
capacity of 40 000 MW and 37 000 MW, respectively. The implementation of construction pro­
grammes has been accompanied from the first by large-scale research programmes on safety. 
Although it can be accepted that the existing power stations are being operated at a satisfactory 
safety level, it must also be acknowledged that the research programmes in that area must be con­
tinued for several reasons: the results are needed by designers and power station operators in 
order to improve the level of protection, not only for workers, the general public and the environ­
ment, but also for the installations themselves, by adopting appropriate preventive and safe­
guard measures. They aim at an improvement of knowledge of significant phenomena and 
mechanisms and allow a quantification of the safety margins (conservatisms) contained in the 
design of commercial NPPs at present in operation or under construction. The results of research 
on safety are indispensable to those responsible for licensing and inspections because they allow 
them to stipulate more accurately the operating limits of the installations and to define the mea­
sures to be taken in the event of an accident. In the present political and social context, research 
on safety plays an important role in helping to make nuclear power acceptable to the public by 
enabling the risks associated with nuclear power production to be evaluated more realistically 
and by facilitating their comparison with the risks inherent in other industrial processes. Finally, 
research on safety provides much-needed support for a Community approach in the field of 
safety criteria and standards for light-water power stations. 
The research programmes underway in the Community and in the industrialized countries which 
are developing the concept of light-water power stations, particularly the United States, have 
evolved considerably in recent years. The accident that occurrred in 1979 in the US power station 
TMI-2 brought to light the importance of the operators' role and behaviour under accident condi­
tions and of sequences of events capable of resulting in severe fuel-element damage. 
Moreover, as better understanding of the problems arising from the typical accidents taken into 
account in the design of reactors was achieved, attention was turned, on the one hand to studying 
abnormal events which, whatever their extent, perturb the normal operation of the installations 
and, on the other hand to looking into much more serious and improbable accidents. These 
activities are described in the proposal for a programme set out below. 
Community research on light-water reactor safety 
Community research activities concerning the technological problems of nuclear safety are con­
ducted under the multiannual programme of the Joint Research Centre, and reactor safety has 
been the most important subject area of that programme since 1973. The current programme 
(1980-1983) has a budget of 172 million ECU and a staff of 716, more than half of whom are 
19.9.83 Official Journal of the European Communities No C 250/9 
involved in work relating to the light-water reactor family. The programme includes theoretical 
and experimental activities to analyze accidents and their consequences and to improve method­
ologies and instrumentation for accident prevention. An important effort is devoted to the 
development and application of new and advanced methodologies to reduce the uncertainties of 
probabilistic risk assessment. 
Analytical and experimental activities are mainly concentrated on the study of the physical phen­
omena dominating loss-of-coolant accidents and in general transients involving also severe deg­
radation of the core. The objective of these studies are to improve and validate the 'safety codes' 
and possibly to gain new information for setting up more efficient operational and emergency 
procedures. 
Two important projects are under way in this field: the out-of-pile LOBI and the in-pile Super-
Sara. 
The main objectives of the LOBI project are: 
— the performance of loss-of-coolant experiments by simulating pipe ruptures of various break 
sizes at three different positions within the 'broken' LOBI loop with the aim of investigating 
the influence of the thermohydraulic behaviour of the individual primary cooling system 
components on the course of a loss-of-coolant accident (blowdown) by measuring all signifi­
cant thermohydraulic quantities, especially those relevant to the core cooling; 
— the application of the experimental results for validating and improving blowdown com­
puter codes and associated theories used for the safety analysis of LWRs. 
The Super-Sara test programme consisted of a series of in-pile experiments (in the ESSOR 
reactor) redefined in 1980 on an international basis. During 1980-1982 the construction of the 
loop, the preparation of the experimental tests and the discussion of the technical content of the 
different phases of the programme with the experts and national delegations were carried out. 
These activities are complemented by two other projects: 
— The project primary system integrity is mainly devoted to developing procedures and calcula­
tion methods which should allow defect detection in structures and a reliable prediction of 
the end of life of reactor components and structures. Among the above, the reactor vessel is 
clearly the most important one and the choice of parameters and boundary conditions in the 
experimental tests and calculations is made accordingly. The JRC is acting as scientific coor­
dinator of the international Programme for the Inspection of Steel Components (PISC) 
sponsored by the CSNI. The JRC is also preparing a new experimental programme on a Vs 
scale pressure vessel: the aim is to set up a systematic approach to the inspection problem. 
— The two main activities included in the reliability and risk evaluation project, accident 
sequence analysis and the European Reliability Data System (ERDS) are strongly related. 
The accident sequence analysis activity is devoted to identifying and modelling class of acci­
dent sequences before and beyond DBA, taking into account, in particular, the timing of 
events and man/system interface. 
The main objective of the ERDS is to implement a centralized data bank system that will 
provide the information needed for risk assessment for LWRs. This bank, or part of it, is 
becoming operational. Data are supplied by different national organizations and the most 
important objective is now to perform systematic analysis of all this information. 
In 1979 a modest shared-cost programme was adopted for the 1979-1983 period. That pro­
gramme on the safety of thermal water reactors had a budget of 6,3 million ECU. Theoreti­
cally, this enabled a 50 % financial contribution to be made to work on the three topics listed 
below, but the contribution was only 37 % in practice owing to the budget limitations 
imposed on research carried out in national bodies or laboratories in Member States: 
No C 250/10 Official Journal of the European Communities 19.9.83 
— studies on separate effects relating to the loss-of-primary-coolant phenomenon, particu­
larly during the core rewetting — reflooding phase (Area A); 
— studies relating to the protection of nuclear installations against gas-cloud explosions 
(AreaB); 
— studies on the atmospheric dispersion of radioactive products released as the result of 
an accident (Area C). 
This programme, which is currently being implemented, will be completed at the end of 
1983. It can already be anticipated that its results, which will be published, will represent an 
important scientific attainment for the Community in respect of the three topics in question. 
The Commission intends to continue and to reinforce its central role in the field of nuclear 
safety by the two types of actions at its disposal, e.g. direct action realized in the laboratories 
of the JRC and by the shared-cost action undertaken in the laboratories of Member States. 
In this view the action programme proposed by the Commission for the development of 
energy by nuclear fission, of which nuclear safety constitutes the principal objective, largely 
takes into account the results of the two programmes in progress mentioned briefly above. 
As to the shared-cost action for the safety of light-water reactors, the second programme pro­
posed for a period of four years, while continuing the research undertaken under the first 
programme, will cover a much wider range of technical subjects. 
Technical field adopted— elaboration of the programme 
As was done in the past for the purpose of defining the first programme on the safety of thermal 
water reactors, the Commission, in 1981 and 1982, consulted the Community's Working Party 
No 2: Safety of light-water reactors — research in order to select the topics to be covered in the 
new expanded research programme. With the aim of going into greater detail in certain fields, the 
Commission, on the advice of Working Party No 2, set up ad hoc subsidiary working parties 
which put forward recommendations that form the basis of the proposals set out below. As 
regards further work to be performed on the three research topics covered by the first programme, 
it was agreed that the recommendations would be made by the three relevant study groups which, 
since approval of the first programme, had been acting on behalf of the ACPM — Safety. These 
three study groups thus put forward recommendations which were taken into consideration in 
this proposal as well. Finally, with regard to certain subjects more directly linked with the current 
direct-action programme of the JRC, the Commission made proposals which were communicated 
to Working Party No 2 for its opinion but which, due to lack of time, did not result in a detailed 
exchange of views within that Working Party. The conclusions and recommendations prepared 
by the various abovementioned specialized study groups are available. 
The research under consideration can be divided into three categories: that concerned with acci­
dent prevention, that is to say research work to develop actual measures to reduce the probability 
of accident occurrence; that which comprises the acquisition of fuller knowledge of accident 
phenomena so that their consequences can be limited; and that which concerns probabilistic 
assessment techniques and methods. However, this classification must not be interpreted too 
rigidly since there are obvious connections between these three categories. It should be noted that 
some of the fields adopted are not specific to light-water nuclear power stations but can also con­
cern other types of nuclear installation; this is the case, in particular, with human factors and 
man-machine interactions, the protection of power stations against explosive gas clouds and, to a 
lesser degree, certain parts of research on other topics that has been proposed. These fields have 
nonetheless been included in the proposal for a programme relating to light-water reactors 
because these reactors, being by far the most numerous, will be the main beneficiaries of the 
results of research in these fields. 
1.1. HUMAN FACTORS AND MAN-MACHINE INTERACTION (Reference: point 
l.A.2.2 of the RAP — research action programme) 
Objectives 
The role of man and the problems of the interface between man and machine have long 
been analyzed and taken into account in non-nuclear, activities in which it was accepted 
that it was economically profitable to be concerned about such problems and to optim-
. 19. 9. 83 Official Journal of the European Communities No C 250/11 
ize man's role as a function of his behavioural characteristics and his capabilities. In 
other processes, such as air transport, human activity was studied from the standpoint of 
safety. Where the operation of nuclear installations is concerned, although the role of 
the operator was not neglected, there was a tendency to believe that properly studied 
and designed automated systems could always cope with dangerous situations as a last 
resort. The TMI-2 accident has made it necessary to revise this approach, and numerous 
debates and studies have since then been undertaken in the Member States, the USA 
and at international level (OECD-NEA, Halden, etc.). 
At Community level, during 1979, 1980 and 1981 the JRC organized several interna­
tional workshops on the behaviour of operators in accident situations, moreover a group 
of specialists has met on several occasions since 1981 under the auspices of Working 
Party No 2: Safety of light-water reactors — Research. The proposal set out below 
amply reflects the conclusions of that group of specialists, taking into account the 
results of the workshops organized by the JRC. The objectives to be attained are clearly 
defined; they are aimed at improving the qualification and the training of operators 
together with the resources at their disposal, particularly in the control rooms. The con­
struction of improved simulators and of control rooms of advanced design is facilitated 
by the technical advances achieved in the fields of components and of data processing, 
but the study of such systems is based more on development than on research and is 
generally conducted by industry in close cooperation with the operators, since industry 
is bound to possess more detailed and more extensive theoretical knowledge of human 
reactions and reliability in "on-routine situations. 
The objectives of the research work proposed here are to a large extent upstream of the 
man-machine interface constituted, for example, by a control room or a simulator. This 
research work should make it possible to improve basic knowledge of operator behav­
iour from the standpoint of mechanisms for information acquisition by taking into 
account factors, circumstances and the environment which influence them. The aspects 
relating to the way in which the work of the teams of operators is organized and to 
means of communication must be studied more systematically. Such theoretical knowl­
edge will then be applied to the definition of the models that can be used for probabil­
istic risk assessment and for the improvement and rationalization of procedures; it will 
also make it possible to direct development along the desired lines and to validate new 
sophisticated diagnostic aid systems. 
This is a new field where Community research is concerned; the corresponding activi­
ties in the Member States are limited and the large scale programme proposed below 
will consolidate and centralize them, and the implementation by the JRC of ERDS 
could facilitate the collection, processing and use of data relating to human reliability 
which are required for the research under consideration. The development at the JRC of 
methods used for probabilistic risk analyses (PRA) will facilitate the application of 
these techniques to human reliability. The shared-cost activities for which provision is 
made below and the direct action project conducted at the JRC will thus be closely 
dovetailed. 
Activities 
The programme is planned to cover three areas wich will logically be dealt with in con­
secutive order. Initial work will thus be focused on the first area. However, this will not 
preclude the concurrent performance of some work in other areas: 
— the collection and analysis of relevant human behaviour data; 
— modelling of man's behaviour; 
— consideration of the suitability of such models for use in probabilistic risk assess­
ments and for design purposes. 
The topics are grouped within these areas. See the Annex for further details and advice 
from the experts. 
No C 250/12 Official Journal of the European Communities 19.9.83 
1.1.1. Collection and analysis of human behaviour data 
Much further work is required, particularly on cognitive processes in real life situations. 
The absence of qualitative and quantitative data on authentic situations is responsible 
for the lack of progress in this area. 
A prerequisite is the creation of an adequate data base: generic error mechanisms can be 
deduced by collecting information on human errors from the nuclear and non-nuclear 
industries. In the case of the latter, many data are already available in the chemical, coal 
and steel industries (reference work within the European Coal and Steel Community): 
— current techniques of human behaviour quantitative data collection. Consideration 
and specification of a homogeneous system and protocol to obtain such informa­
tion. Possible development of automatic, on-line data collection equipment, 
— existing information in past records of NPP, 
— applicability of the large amount of information obtained in non-nuclear activities, 
— techniques to extend the applicability of objective data to situations other than 
those in which it was collected, 
— setting up of particular test series to obtain further specific information, 
— current data taxonomy and classification schemes, 
— methodology for event analysis, taking into account human and equipment failure 
interaction, 
— relationship of human reliability factors and plant whole life cycle, 
— environmental, intellectual and emotional effects on the human individual, 
— organizational and communication problems in the whole life cycle of nuclear 
power plants. 
1.1.2. The modelling of man's behaviour 
Particular attention is given to reliability aspects of such behavioural models. 
— Study and critical assessment of existing models. 
— Consideration and application of complex mathematical modelling techniques to 
develop new models. 
— Model comparison exercises, by means of benchmarks, controlled experiments, etc. 
on both real plants and simulators. 
1.1.3. The considerations of the suitability of such models in PRA and for design purposes 
Two main areas of use are foreseen for such models: 
— Incorporation in and influence on probabilistic risk analysis. (PRA) studies and 
their application to risk management and reliability assurance. 
— The feedback of relevant information for design purposes. This has many facets, 
including control room design and information flow, operating procedures, 
development and validation of advanced support facilities for systems operation. 
This is a vast area and cannot all be included within this programme. Therefore a step 
by step approach will be followed. 
19.9.83 Official Journal of the European Communities No C 250/13 
Programme requirements 
— Close liaison with the JRC activities and other relevant research programmes of 
DGVandDGIII. 
— Close liaison with laboratories with relevant experience (including non-nuclear). 
Participation with Halden is particularly recommended as well as the US EPRI and 
NRC human factor programmes and the Scandinavian programmes. (Note the rel­
evant proposed COST programme on Systems of socio-technologies and industrial 
safety.) It is essential to have information on activities in all laboratories who are 
carrying out relevant work within the European Communities. 
— Close working relationships with various utilities. Regular contact activity involv­
ing human factors experts and operators is essential and needs to be encouraged by 
positive proposed actions. 
— Much information exists on methods of human factors assessments in non-nuclear 
hazardous industries. There is a need to investigate the feasibility of the transfer of 
such techniques to nuclear activities. 
Community contribution 
Community programme of investigations to be implemented on the basis of research 
contracts: 
funds necessary: 3 600 000 ECU. 
1.2. PROTECTION OF NUCLEAR INSTALLATIONS AGAINST GAS-CLOUD 
EXPLOSIONS (Reference: point l.A.2.6 of the RAP) 
Objectives 
The protection of nuclear power stations and installations against external explosions of 
accidental origin, in this case clouds of heavy gas (hydrocarbons) released in the vicinity 
of power stations, for example after a transport accident, has been receiving an increas­
ing amount of attention for several years. This also applies to various types of non-
nuclear installations (platforms, gas terminals, storage areas, etc.). This protection may 
involve regulatory measures concerning the siting of dangerous installations and the 
transport of inflammable substances together with preventive measures when power sta­
tions and their auxiliary facilities are being designed. 
The phenomenology of such unconfined external explosions is very complex and still 
poorly understood. It is for this reason that theoretical and experimental research into 
various aspects of such explosions (formation and migration of the explosive cloud, for­
mation and propagation of a pressure wave following deflagration/detonation, interac­
tions of such waves, etc.) have proved to be indispensable. Such studies have been 
undertaken on a modest scale in the context of Area B of the first shared-cost research 
programme (1979-1983) and have made it possible to consolidate and sometimes stimu­
late a significant part of the work conducted in the Member States. This action must 
now be continued and supplemented since both the paucity of the available resources 
and the relative novelty of the field have made it necessary to concentrate research work 
on a number of priority aspects: intercomparison of deterministic models and valida­
tion by means of a prototype experiment of instantaneous releases (2 000 m3) under iso­
thermal conditions at Thorney Island in respect of the dispersion aspect; development 
of simple codes describing the formation and propagation of pressure waves, but not 
including a detailed prediction of the combustion, and validation by means of 
medium-scale experimental surveys (balloon explosions) in respect of the explosion 
aspect; analytical effort describing these phenomena in the presence of structures and 
validations by means of measurements on mock-ups or at the water table in respect of 
the shock-wave propagation and interaction aspect. 
No C 250/14 Official Journal of the European Communities 19. 9. 83 
The Commission, in close liaison with the ACPM study group responsible for monitor­
ing the work relating to Area B of the first programme, considered what further work 
should be done along those lines under a second programme, taking into account at the 
same time the results acquired, the aspects that were not covered and general develop­
ments in the field and in certain experimental techniques. The conclusions drawn from 
that study were taken into account in the list proposed below. 
Activities 
These can be divided into five categories; the first and the last were not dealt with in the 
first shared-cost programme. 
1.2.1. Source term 
By this is meant the initial circumstances and parameters of the accidental release 
before formation and migration of the explosive cloud. Particular attention will be paid 
to the phenomenology and physics of the initial phase of cloud formation, in particular 
from initially liquefied gas, and to the evaporation mechanisms. 
1.2.2. Dispersion 
The Thorney Island releases have already provided a valuable set of reference cases. A 
detailed analysis of these data and a comparison between them and the predictions 
derived from various models and codes should be carried out under the second pro­
gramme. Additional releases, for example non-isothermal releases, could supplement 
the results of the initial measurement surveys. 
However, one implicit objective of these experimental surveys is also to validate the use 
of wind tunnels and water tunnels for reduced-scale simulation of the migration of lay­
ers of heavy gas at ground level or at the surface of the water in the presence of ob­
stacles or heat pockets. The feasibility of such simulations was demonstrated during the 
first programme and very sophisticated analytical methods have become operational. 
For this reason, a major effort will have to be made in a field in which it should be 
possible to perform the study of realistic cases or geometries at reduced cost. 
Moreover, all the currently,available predictive models are deterministic and do not 
take account of the considerable intrinsic variability of the dispersion or of the very irre­
gular structure of the clouds, which may markedly influence the explosion phenome­
non. Particular attention will thus be paid to the preparation of stochastic models and to 
detailed studies and validations, for example in wind tunnels. This point is added to 
that referred to in Section 1.7 which concerns the atmospheric dispersion of radioactive 
products. 
1.2.3. Combustion and formation of a pressure wave 
Much still remains to be done in this complex and poorly understood sector. What is 
mainly lacking is theory, models and satisfactory codes for combustion/flame propaga­
tion in an unconfined or partially confined cloud; the pressure wave created depends 
very critically on how this propagation takes place. All that is available is fragmentary 
information and results which were generally obtained at laboratory level on flame 
acceleration mechanisms and the possible transition from deflagration to detonation. 
The limited effort made in the first programme will thus have to be continued and 
intensified, and should also incorporate other aspects: influence of irregular cloud struc­
ture, of partial confinements or obstacles and of the topography; or, in order to achieve 
more realistic conditions: influence of the shape of the ignition point, of the presence of 
aerosols, etc. In the main, it would involve studies and simulations of partial effects. A 
concerted series of tests in the field, on a large scale, could be undertaken during the 
19. 9. 83 Official Journal of the European Communities No C 250/15 
implementation of the programme or, if necessary, be replaced by participation in an 
international experiment of that type. It goes without saying that development of 
predictive models and codes will accompany these different activities. 
1.2.4. Propagation and interactions of pressure waves 
The limited effort made during the first programme will have to be continued. At least 
one elaborate code (isentropic assumption) will be developed. If required, it will be 
coupled with a description of the combustion and reaction of structures. The tests in the 
field will be continued in order to deal with the case of multiple sources, of propagation 
above the sea and of the influence of topography or in order to validate certain aspects 
of the predictive codes. 
1.2.5. Reaction of structures 
This fundamental aspect, which is quite distinct from the preceding ones, was not dealt 
with in the first programme. The numerous available data on the reaction of conven­
tional buildings to explosions and the damage suffered by them can be extrapolated 
only with difficulty to the reinforced structures of nuclear power stations, and it is for 
this reason that tests on structural mock-ups or elements and an examination of the con­
sequences of simultaneous cases of damage for the safety of an installation will have to 
be underaken in close liaison with the JRC, which possesses specific skills in this field. 
Community contribution 
Joint programme of investigations to be implemented on the basis of research contracts: 
funds necessary: 4 200 000 ECU. 
1.3. MECHANICAL AND MATERIAL PROBLEMS WITH STEEL COMPONENTS 
IN LIGHT-WATER REACTORS (Reference: point l.A.2.3 of the RAP) 
Objectives 
The integrity of steel components in light-water power stations, particularly those in pri­
mary circuits, is essential for the safety of such installations. Major research pro­
grammes have so far been devoted to this area. Their results have been used to improve 
manufacturing processess and quality-control techniques. Attention was then turned to 
the development and application of systematic and periodic controls throughout the 
liefetime of the installations. Early fault detection and the monitoring of fault develop­
ment were justified, in particular, by the phenomena of radiation-induced steel embrit-
tlement and of steel fatigue caused by the thermal cycles. Recently, this trend towards 
emphasizing inspections received additional impetus from the discovery of faults 
beneath coatings, from thermal shock problems, etc. 
International cooperation in respect of the mechanics and materials of steel compo­
nents in light-water power stations has existed for a long time and is centralized in a 
working party under the joint responsibility of the CSNI (OECD-NEA) and the Com­
mission. From 1976 to 1980, an international programme for the detection of already 
existing faults in steel pieces which had been circulated among a large number of spe­
cialized laboratories was implemented (PISC I). The JRC played an important role in 
that programme by carrying out the destructive analysis of the test specimens and the 
processing and application of the results obtained by the laboratories. The experience 
derived from that programme resulted in the setting up of a second programme, 
PISC II, with special emphasis on the validation of non-destructive fault detection, 
location and evaluation techniques and on the definition of efficient and reliable 
inspection procedures; here also the JRC is responsible for central coordination. 
During the same period (1976-1980) a study aimed at the evaluation of possibilities of 
reactor pressure vessel rupture was promoted jointly by JRC, CEA and Framatome with 
No C 250/16 Official Journal of the European Communities 19.9.83 
the participation of European industries constructing pressure vessels. The study, based 
on the experience of fabrication and inspection of about 15 pressure vessels, allowed 
identification of parametric relations between loading conditions, purity of material and 
performance of systems for inspection. For the programme 1984-1987 at the JRC the 
continuation of an important research effort is foreseen in the field of an early detection 
of failures and in the development of methodologies for predicting the life period for 
primary circuit components. 
The Commission deemed it advisable to propose research relating to the mechanical 
and materials problems affecting the steel components in the primary circuits of light-
water reactors as part of a shared-cost project to support and supplement the activities 
conducted at the JRC in that sector. The shared-cost project provides a well adapted 
framework in view of the close international cooperation which was established during 
the PISC programmes and within the theoretical-experimental programme on pressure 
vessel models at reduced scale and safe end models at scale 1 commenced at the JRC. 
The national laboratories participating in PISC II will be in a position to assume res­
ponsibility for the shared-cost projects proposed by the PISC group itself and supported 
by the international working parties as an extension of what is currently being done on 
the basis of the JRC programme. The projects under consideration will be centred on 
the following points: where evaluation of the effectiveness and reliability of non­
destructive testing methods is concerned, parametric studies of the sensitivity of these 
methods with different factors will be carried out and programmes for their comparison 
and validation (benchmark) will be implemented. Where fracture mechanics is con­
cerned, work on residual and thermal stresses will extend the activities under way at the 
JRC and make them profitable. 
Finally, a further field which is relatively independent of the preceding ones concerns 
the mechanical problems caused by accidents of external origin, in particular by earth­
quakes: this will initially involve taking stock of the available knowledge, comparing 
the various methods of investigation and setting up a programme for validating codes 
applied to an elementary case. It should be noted that, although most of the research 
proposed below is directly related to light-water power stations, it is also of general 
interest. 
Activities 
1.3.1. Evaluation of the effectiveness of NDT techniques 
Several areas exist where a very efficient contribution could be given to the Programme 
of inspection of steel components (PISC II). These areas are those called 'Parametric 
Studies' which started on a very limited programme end of 1982 as part of the CCR 
programme but which should be extended along the lines proposed by international 
experts groups. The actions required are typically 'cost sharing' type actions as it is 
necessary to involve equipment, know-how and skilled manpower of developers and 
users of NDT techniques and inspection procedures for nuclear industry in general and 
reactor components in particular. 
Six groups of parameters which require parallel efforts are identified: 
— the effect of the defects characteristics, geometry and position on detection, loca­
tion and sizing; 
— the effect of the equipment characteristics on defect detection, location and sizing; 
— the effect of cladding characteristics on defect sizing; 
19. 9. 83 Official Journal of the European Communities No C 250/17 
— the parameters involved in the use of electromagnetic techniques for near surface 
detection; 
— the effect of residual stresses on defect detection, location and sizing; 
— the evaluation of results including signal treatment and visualization. 
The national organizations likely to execute actions on these parametric studies will 
introduce present standard inspection procedures as well as advanced NDT techniques. 
The considered material is generally pressure vessel steel and defects are realistic ones, 
artificially introduced or induced to allow all possible comparative benchmark exercise. 
As a consequence of the results of the evaluation in the PISC exercise, validation proce­
dures and demonstrations will be necessary. Information programmes, support and 
coordination actions could be necessary. 
Actions such as: 
— information and sensitization of codes, standards and regulatory bodies concerned 
with ISI (in-service inspection), 
— collection/creation of validation samples and structures for validation laboratories, 
— coordination for a better use of the material available in the European Community 
are typically the competence of CEC. 
1.3.2. Detection, monitoring and sizing of service induced defects in loaded structures and assess­
ment of residual life 
The programme aims in general at assessing the capability of combined NDT tech­
niques to predict by periodic or continuous monitoring the residual life of loaded struc­
tures. 
The JRC started in the years 1982-83, and is planning to develop in the years 1984-
1987, a theoretical and experimental programme on scale Vs PWR pressure vessels and 
on a scale Vi specimen of piping and specifically safe-end. The experimental facility 
allows programmed loading histories, automatic internal US scanning, computerized 
data acquisition and analysis for US, acoustic emission, strain gauges, laser holography. 
Supplementing the work performed by JRC, the cost-sharing programme should allow 
specialist teams from certain laboratories in Member States to realize runs of measure­
ments on pressure vessels and installations of the JRC as well as preliminary work 
mainly in support of: 
— defect sizing and defect danger assessment by acoustic emission; 
— comparative evaluation of US techniques and US signal analysis software on ser­
vice induced defects; 
— correlation of various NDT techniques measurements (US, acoustic emission, RX); 
— testing and comparison with FEM codes of experimental stress analysis techniques 
(e.g. laser interferometry, thermal emission); 
— benchmark exercise on elasto-plastic fracture mechanics in tri-dimensional loaded 
structures (nozzle). 
No C 250/18 Official Journal of the European Communities 19.9.83 
1.3.3. Fracture mechanics 
Residual and thermal stresses 
Investigation of existing problems in the following areas: 
— measurements of residual stress distribution in, and adjacent to, welds; 
— residual stresses associated with cladding and underclad cracking; 
— influence of fabrication and heat treatment practices on residual stresses and a 
comparison of codes and specifications; 
— effect of hydro-test on residual stresses; 
— the influence of residual stresses on ductile-brittle transition behaviour. 
Numerical analysis of elasto-plastic fracture mechanics problems 
It is proposed to perform a calculational round robin to arrive at a better understanding 
of the most suitable techniques to be applied, in particular with respect to the following 
problems: 
— type of modelling in the vicinity of the crack tip; 
— optimum choice of the iteration parameters in the elastoplastic analysis; 
— treatment of stable crack growth. 
In addition it is suggested to compare the results of the numerical analysis in part or 
totally with results obtained experimentally by defining limits and peculiarity of the 
analysis used. 
1.3.4. Mechanical problems following shock waves caused by accidents of external origin (in 
particular by seismic ground motions) 
The investigation of the limits of applicability of the different methods applied to calcu­
late the consequences of seismic loads (and other loads originating from external acci­
dents) on internal structures (including reactor components and pipework) is suggested 
in three steps: 
— preparation of an inventory of existing analysis and testing methods; 
— comparison of different approaches to define convergencies and divergencies; 
— development of improved methods. 
Community contribution 
Joint programme of investigations to be implemented on the basis of research contracts: 
funds necessary: 4 800 000 ECU. 
1.4. THERMOHYDRAULICS AND SEVERE DEGRADATION UNDERGONE BY 
FUEL DURING LOSS-OF-PRIMARY-COOLANT ACCIDENTS (Reference: points 
l.A.2.4 and l.A.2.5 of the RAP) 
Objectives 
The accident that occurred at the US power station TMI-2 was one of the main factors 
which led to a reorientation of research programmes in these fields after 1979. Although 
19.9.83 Official Journal of the European Communities No C 250/19 
work had previously been concentrated on the detailed study of the loss-of-primary-
coolant accident of the 'large leak' type and it was considered that all the other situa­
tions were less harmful than that major accident, considerable importance was hence­
forth attached to other scenarios capable of resulting in large-scale fuel damage and 
likely to give rise to unrecoverable situations. The case of small leaks affecting the pri­
mary circuit was only the most characteristic case of these types of scenario which are 
studied in detail because of their consequences in respect of the installation, the envi­
ronment, the workers and the general public. At Community level, as part of the direct-
action programme implemented at the JRC, the out-of-pile LOBI loop system initially 
intended for the study of transients relating to a large-leak and intermediate-leak loss-
of-primary-coolant accident during the depressurization phase was modified to enable 
'small-leak' transients to be created. 
The LOCA-ECCS aspect (Area A) of the current shared-cost research programme covers 
the performance of out-of-pile experiments on separate effects relating to thermo-
hydraulics and to heat exchange in the core rewetting-reflooding phase. It involves 
research of general interest into one- and two-dimensional geometries and two-phase 
conditions. The effects caused by deformed channel geometries simulating the swelling 
of cladding and by the rod composition were taken into account in several experiments, 
and it may be said that this part of the programme was at the dividing line between 
thermohydraulics and thermomechanics. 
In 1982, the Commission in close contact with the ACPM study group responsible for 
monitoring the implementation of the first programme in this field performed a study in 
order to determine what further work should be done along those lines after 1983. The 
Commission considered that it was necessary to continue the fundamental work under­
taken in the field of thermohydraulics and heat transfer under two-phase conditions. 
This research should make it possible to establish correlations and models directly 
usable in the preparation of realistic codes (best-estimate). 
Emphasis will have to be placed on 'small-leak' loss-of-coolant situations and on the 
development of measurement techniques appropriate to two-phase conditions. The 
effects arising from reduced-scale simulation of the phenomena will have to be brought 
to light in such a way as to validate the application of the results obtained in real reactor 
cases. The study group also recommended progressive continuation of research work on 
situations leading to core degradation, particularly with regard to the cooling of 
deformed fuel and even including the rewetting of the debris produced after cladding 
meltdown. This specific area of research on damaged fuel was examined in greater 
detail in 1981 and 1982 by a group of specialists convened under the aegis of Working 
Party No 2: Safety of light-water reactors — Research. The group took into considera­
tion the numerous projects under way or planned in the Member States of the Com­
munity and outside the Community, particularly in the USA and at international level 
within the framework of the CSNI (OECD-NEA). It made recommendations which the 
Commission took into account in proposing a number of experiences of the kind of 
separate-effects experiments referring to two categories of situations: those involving 
degraded core conditions capable of resulting in the formation of debris (rubble bed) 
and even in localized meltdown of fuel pellets, in which case the situation is recoverable 
and its evolution can be stabilized by re-establishing some method or other of cooling. 
It is in this field that most of the work will have to be concentrated, more particularly 
with regard to the formation and characterization of debris as a function of the various 
cooling systems, to cooling under single-phase and two-phase conditions in degraded 
cores and to the reactions and physical properties of the materials at increasingly higher 
temperatures. Another category of situation is the unrecoverable one and involves core 
meltdown and melt-through of the support plates within the pressure vessel together with 
the passage of corium across the vessel and the containment. This will have to be taken 
into consideration, but at a lower level Of priority than the recoverable situations. In the 
initial stage, the attack on the basemat by the core should be studied. However, the ini­
tial results of the examination of the fuel and the internal parts of the TMI-2 pressure 
vessel will have to be taken into account in the more detailed definition of the work to 
be undertaken. 
The results of the separate-effect experiments have to be confronted, correlated and val­
idated by large-scale in-pile experiments of the integral type simulating or reproducing 
No C 250/20 Official Journal of the European Communities 19. 9. 83 
all the phenomena appearing in accident scenarios. The relations existing between fun­
damental phenomena studied separately and global phenomena from by in-pile experi­
ments can only be brought to evidence by thorough theoretical analysis and modelling. 
The Commission, convinced of the necessity to touch the problems connected with 
damage of fuel under severe accident conditions on the three levels which constitute 
separate-effect experiments, global in-pile experiences and theoretical aspects of 
analysis and modelling proposed, in the framework of direct actions, the creation of a 
group of specialists for analysis and modelling. 
This group has to be informed of the results of significant in-pile experiences concern­
ing severe damage of fuel. Following the discontinuation of the Super-Sara project, 
the French Phebus programme constitutes, on a European level, the programme most 
likely to give the information required, but likewise the programmes in the US like PBF, 
in Canada like NRU or international like LOFT have to be taken into account. The 
availability of the results of the Phebus programme logically has to be searched for 
within the framework of the shared-cost action and negotiations with third countries in 
order to get to know the results of the other programmes mentioned above have to be 
envisaged according to the lines proposed under point 4 below. It is to be seen that in 
this field shared-cost action and direct action will consequently be closely linked and 
complementary to each other. 
Generally speaking, in proposing the above research work, the Commission's objective 
is to promote a certain number of tasks (separate effects) and also to allow availability 
on the Community level of results of important in-pile programmes in progress or envis­
aged in certain Member States or third countries using the analyzing capacities of the 
JRC in this field. 
Activities 
1.4.1. Loss-of-coolant accident — thermal — hydraulics (fuel temperature < 1 200 ° C) 
follow-up of area A of the first programme (1979-1983) 
(a) basic studies based on small-scale experimental work with detailed evaluation lead­
ing to models as input to best-estimate codes and helping the interpretation of 
large-scale experiments such as: 
— transition boiling at the minimum film boiling point; 
— basic heat transfer in bundles and simple geometries relevant to small break 
LOCA in case of high-pressure low-mass flow-rate conditions; 
— condensation effects related to injection of ECCS water; 
— study on the separation of phases in a junction in a horizontal pipe with a 
two-phase low-velocity flow, especially in case of high pressure; 
— two-phase flow patterns, interfacial areas, relative phase velocities expecially 
in case of large diameters D > 200 mm. 
(b) Evaluation/modelling work on available experimental data coming from large-
scale separate- and system-effect experiments with the aim of code validation and 
assessment and of data scaling for transposition to reactor scale conditions. In this 
field the work will be closely coordinated with the activities undertaken within the 
direct action programme and the specific competence of the JRC could be used in 
support of actions at shared cost comprising in particular: 
19.9.83 Official Journal of the European Communities No C 250/21 
— counter current reflooding restrictions and ECC bypass (core region); 
— counter current flooding restriction in pipes; 
— single and two-phase natural convection phenomenon (SB-LOCA configura­
tion); 
— modelling of U-tube steam-generator functioning in secondary and primary 
side two-phase flow condition. 
1.4.2. Degraded core and severe fuel damage studies — recoverable situations 
This situation involving debris formation and eventually some local melting is likely to 
be recoverable with partial reinstatement of some cooling systems and the appropriate 
actions. (Note connections with LMFBR phenomenology.) 
Studies in this domain will be aimed at defining boundary situations within which the 
recovery of a long-term coolable state could be guaranteed. Knowledge of coolability 
limits is necessary in order to define the actions required and to establish the recovery 
procedures. 
(a) Material reactions — physical/chemical processes and kinetics 
— clad oxidation 
high temperature oxidation of zircaloy and stainless steel in various steam/ 
hydrogen mixtures 
influence of pressure on oxidation kinetics 
influence of endothermic reactions on melting timescale 
— clad/fuel liquefaction, candling and relocation 
low temperature U02/Zr02 eutectic formation 
melt progression mechanisms 
— materials interactions 
release of absorber alloy as a function of temperature and its influence on sub­
channel blockage. 
(b) Debris bed formation and cooling 
— debris formation under various cooling regimes 
the effect of quenching on fuel assembly integrity as a function of heating rate, 
chemical environment and quenching modes 
physical characteristics of debris resulting from quenching 
— thermal-hydraulics of recooling a degraded core 
Single- and two-phase flow hydraulics in debris beds. Permeability, channel­
ing, fluidization. Characterization of debris; validity of uniform particle 
studies and simple geometries. Consideration of more complex beds 
Heat transfer in a particle bed cooled by two-phase mixture; quenching of an 
overheated bed by top/bottom injection of water and determination of limit­
ing factors, e.g. steam binding. 
No C 250/22 Official Journal of the European Communities 19.9.83 
(c) Mathematical modelling 
To some extent continuous theoretical modelling work is envisaged on advanced 
severe accident computer codes. This will include benchmark exercises using a 
reference (non-specific) NPP design. A workshop could be held to discuss results 
and produce recommendations. 
1.4.3. Degraded core and severe fuel damage studies — unrecoverable situations 
This area is entered when the preceding coolability limits are passed. Coolant injection, 
if any, would not be efficient enough to prevent bulk melting and fuel collapse on the 
support plate. 
Progression of the accident may involve support plate melt-through, pressure vessel fail­
ure and debris interactions in the containment. 
Work proposed here is on problems relevant to basemat attack. Work should be ini­
tiated to specify an experimental facility, possibly based on one already existing. 
Programme will be specified in detail when TMI-2 is examined in 1983-1984. 
Specific problems are: 
— coolability of debris under water in the reactor cavity taking into account realistic 
debris geometry, 
— likelihood and effects of steam explosion after arrival of debris in the reactor 
cavity, 
— the hydrodynamics of cooling of debris by inundation by water from above (e.g. 
discharge of accumulators) and partition of energy between steam generation and 
concrete, 
— the attack of molten core debris on the floor and walls of the cavity, including 
assessment of the penetration envelope, 
— effect of chemical interactions between gas evolved and the steel phase in the 
molten debris on the composition of the gases reaching the secondary containment 
and on the physical and chemical constitution of the molten debris, 
— long-term cooling and ultimate form of the core debris after penetration into the 
containment basemat. 
1.4.4. In-pile experiences — behaviour of fuel under severe accident conditions 
Participation of the Commission in the Phebus programme and in other international 
programmes or programmes undertaken by third countries: LOFT, PBF, NRU, 
EPRI... 
Community contribution 
Community programme of investigations to be implemented on the basis of research 
contracts: 
funds necessary: 14 200 000 ECU. 
19.9.83 Official Journal of the European Communities No C 250/23 
1.5. PROBLEMS CONCERNING THE DISTRIBUTION, COMBUSTION AND 
MONITORING OF HYDROGEN PRODUCED IN THE EVENT OF AN 
ACCIDENT (Reference: point l.A.2.6 of the RAP) 
Objectives 
From the standpoint of the study of a loss-of-primary-coolant accident followed by 
proper operation of the emergency core-cooling system, the hydrogen production phe­
nomena resulting both from the zirconium-steam reaction and from radiolysis are slow 
enough to allow the dilution or recombination systems to operate before the inflam­
mability limit is reached. The regulations laid down in this respect are confined in gen­
eral to establishing a maximum percentage of cladding oxidation in order to limit the 
quantity of hydrogen produced during the design basis accident. 
When accident scenarios are studied which lead to overheating of the exposed fuel ele­
ments above 1 200° C likely to give rise to increasingly severe core damage, it can be 
seen that the hydrogen production phenomenon occurs more rapidly and on a larger 
scale. In this case, the resultant risks to the integrity of the containment and the safe­
guards systems inside it must be taken into account and appropriate measures must be 
considered. This emerged from the TMI-2 accident, after which ambitious research pro­
grammes were put in hand, particularly in the USA by EPRI and Sandia. In 1979, the 
Commission decided to set up a study group of specialists under the responsibility of its 
Working Party No 2: Safety of light-water reactors — Research, and this study group 
drew up a report in August 1981 on the state of the art in this field. The conclusions 
reached in the report and the subsequent discussions which took place within that study 
group pointed out the need to undertake research at Community level, account being 
taken of the current research programmes in the Community and the United States. The 
fields that were chosen are: hydrogen distribution within the containment, which gives 
rise to major problems with regard to the evaluation and limitation of risks, combustion 
and a study of the likelihood of detonations of hydrogen-air-steam mixtures and moni­
toring, particularly with regard to measurements of hydrogen, oxygen and steam con­
centrations. 
Since the aspects relating to hydrogen production have already been dealt with in sub­
section 1.4.2 above, priority will have to be given to the study of distribution and com­
bustion, but the field of monitoring should not be neglected. The shared-cost project is 
well adapted to this type of work, which involves several scientific disciplines, including 
a number in which the JRC is not specifically competent. Links with other research top­
ics included in this proposal are close, particularly those projects concerning degraded 
core conditions and research on the source term. Although the problems associated with 
hydrogen combustion are different because of the density of the gases in question, there 
are links between that field and the research proposed in subsection 1.4.2 on explo­
sions of gas clouds, and appropriate contacts will have to be established between the 
laboratories dealing with these research areas. 
Activities 
In this field, close cooperation with the US programmes EPRI and Sandia is essen­
tial. 
The overall proposed programme commences with theoretical and experimental studies 
of gas distribution in realistic containments. Possibilities of combustion are then investi­
gated with the aim of achieving satisfactory technical control of the gases produced 
during an accident. 
1.5.1. Distribution of hydrogen and other gases 
Questions of distribution in post LOCA containment atmosphere may dominate tech­
nical control and risk assessment. 
No C 250/24 Official Journal of the European Communities 19.9.83 
Knowledge of the distribution of atmospheres within the compartments of compart-
mented containments requires analytical and experimental studies on hydrogen, oxy­
gen, steam, nitrogen distribution within subcompartments. Investigations will start with 
basic studies (for example on separation of effects) through to integral tests. Prior to 
those tests analytical studies should be performed for optimization of the whole pro­
gramme on distribution. Of particular relevance to the early stages of an accident is the 
separation of H2 and O-L by physical processes. 
Investigation is necessary on the accuracy of theoretical analyses and practical concen­
tration measurements in accident conditions. 
It is necessary to develop a rugged and reliable method for measuring hydrogen and 
oxygen concentrations. 
For advanced degraded core situations and complete core melt conditions, more code 
development, comparison and experimental verification is necessary. 
The distribution aspects of all mitigation procedures need to be carefully analyzed to 
assess the benefits of these procedures (see technical control). 
1.5.2. Combustion processes 
Bearing in mind current work, several areas could benefit from investigation by Euro­
pean laboratories. 
Whereas the flammability and explosion limits are well known for binary hydrogen-air 
systems, data for ternary and multiple component systems, when steam, inert gas or 
aerosols are present, are very sparse. Reliable data for the explosion limits for the hydro­
gen-air-steam system are required, particularly for realistic, compartmentalized contain­
ments. 
Additional to the aspects of distribution effects on combustion phenomena, special 
attention should be paid to physical conditions prior to and after a possible ignition. 
This needs combined analytical and experimental work. If possible, experiments should 
provide a reliable understanding to predict reaction kinetics of combustion, especially 
the possibility of DDT (deflagration of detonation transition) under real conditions. 
This will include the influence of generated pressure waves and structure interactions. 
This should then be integrated into H2 system codes to study sample scenarios. 
1.5.3. Technical control 
Several different procedures and principles are currently under discussion and evalua­
tion and some of these are due to be tested in the USA programme. 
Firm recommendations on specific techniques cannot yet be made until each procedure 
is comprehensively analyzed, tested and results made available. Complementary Euro­
pean work to the USA programme in this area (especially with distribution aspects) is 
strongly recommended. This work will be performed in the latter part of the proposed 
programme. 
Community contribution 
Joint programme of investigations to be implemented on the basis of research contracts: 
funds necessary: 3 600 000 ECU. 
19. 9. 83 Official Journal of the European Communities No C 250/25 
1.6. FISSION PRODUCTS SOURCE TERM UNDER SEVERE ACCIDENT CON­
DITIONS (Reference: point l.A.2.5 of the RAP) 
Objectives 
Evaluation of the source term requires knowledge of the inventory of fission products 
that exists potentially in a nuclear power station at the moment of, and during the var­
ious phases of, an accident and knowledge of the fission-product release and transport 
process in the primary circuit, then in the containment and finally to the environment. 
This evaluation must be a qualitative one: in other words, it must stipulate the various 
physical and chemical forms in which the fission products are released and develop, 
and it must be quantitative. Knowledge of the source term plays a central role in a real­
istic risk assessment pertaining to the accidents under consideration and makes it pos­
sible to define the countermeasures and safeguards systems which must be planned in 
order to cope with them. 
For the purposes of regulations, the hypotheses taken into account in respect of the 
source term for assessing the radiological consequences of reference accidents are very 
largely conservative, particularly with regard to the releases of solid and semi-volatile 
fission products; only the discharges of noble gases are realistically evaluated. After 
publication of the report WASH 1400, American specialists such as Levenson and Rahn 
drew attention to the pessimisms of the hypotheses and the models used and to the role 
played by water in the retention of iodine and caesium. Their point of view was sup­
ported by the first observations made on the occasion of the TMI-2 accident, after 
which major research programmes were decided on in the USA by NRC and EPRI and 
in Europe. Those programmes concern both the chemistry of certain fission products 
and the formation and behaviour of dense aerosols. In the field of international cooper­
ation, the Swedes proposed that the Marviken installation be used to implement an 
ambitious research programme on aerosols. That programme will start in 1984 and sev­
eral Member States have already announced their intention of participating in it. 
Furthermore, the CSNI (OECD-NEA) specialized working party prepared a report in 
1979 on the state of the art with regard to nuclear aerosols. That report has been revised 
and adapted to light-water power stations, since the first version was largely devoted to 
fast sodium-cooled reactors. In 1982, Community Working Party No 2: Safety of light-
water reactors — Research considered that a wide range of problems still remained to be 
solved in respect of the source term and that it would be desirable to consider research 
projects as part of a second shared-cost research programme. 
In July 1982, the Commission convened a small group of specialists which reviewed the 
activities conducted or planned both within and outside the Community. They recom­
mended that work of a fundamental nature be performed which would be limited to the 
study of separate effects and would not require large-scale facilities. The research work 
will concern the process of fission-product release from fuel at high temperatures, the 
problems of iodine chemistry in an aqueous environment, the transport of fission prod­
ucts in the primary circuit and, in particular, numerous investigations into the transport 
and behaviour of fission products and aerosols in the containment. Preparatory studies 
will have to be undertaken by the Commission in 1983 in order to define more precisely 
a certain number of potential topics and to take into account the results of experiments 
currently under way in the USA and Europe. This involves the need for a certain 
amount of flexibility in carrying out the projects proposed here; it will, in particular, be 
advisable to make provision for calls for tenders at several times. 
The results of this research will make it possible to improve the modelling of the various 
phenomena which has so far been used in the specialized codes. In addition, the 
research proposed here will make it possible to achieve a better definition of the initial 
data required for the study of the problems associated with the atmospheric dispersion 
of radioactive discharges which are dealt with under heading 1.7. 
No C 250/26 Official Journal of the European Communities 19.9.83 
Activities 
1.6.1. Fission product release from fuel 
— Unambiguous identification of the chemical form of iodine and other fission prod­
ucts at the time of release from fuel at elevated temperatures. Because of the com­
plex chemical reactions that are possible these measurements should be done with 
actual irradiated fuel in a steam/hydrogen environment. 
— Release rates of semi-volatile fission products from irradiated fuel 1 200-2 100 °C. 
1.6.2. Chemistry of fission products 
— High temperature thermochemical data and phase relationships for many fission 
product species, plus reactions with reactor materials, with other fission products 
and with vapour constituents at high temperature. Also the physical and chemical 
effects of hydrogen burns. 
— Aqueous iodine chemistry, particularly at elevated temperatures (< 300 °C), on the 
kinetics of known reactions and stability of suspected intermediate species. Parti­
cular study of organic iodide formation at low concentrations. Possible develop­
ment of a code. 
1.6.3. Fission product transport in primary coolant system to containment 
— Critical review of models omitted from Trap-Melt, with development of these as 
necessary and as foreseen as evolving in the future. Code comparison exercise 
around 1985. 
— Experimental research on gravitational coagulation processes. 
— The following areas should be borne in mind as possibly requiring some work: 
Effect of small steam explosions within the PCS. 
Aerosols produced by fragmentation forced out under high pressure from the PCS. 
The significance of possible quench tank attenuation. 
The possible existence of very dense aerosols in certain accident sequences and 
their subsequent stability. 
1.6.4. Fission product transport and aerosol behaviour in containment (Note the past and current 
LMFBR relevant programmes) 
— Condensation processes, influence of noble gases decay heat, development and 
coupling of thermal hydraulic and aerosol codes, possible development of a com­
bined code. 
— Significance and modelling of diffusiophoretic deposition processes. 
— Development of coupled models including natural and engineered removal 
mechanisms. Code comparison and benchmark exercise. 
— Experimental validation of agglomeration processes, particularly gravitational 
coagulation. 
— Study of possible resuspension processes from walls and sump at various stages of 
the accident. 
— Attenuation through leak paths. Scoping studies to define attenuation for certain 
types of leak path and what parameters are most significant. 
19. 9. 83 Official Journal of the European Communities No C 250/27 
Community contribution 
Joint programme of investigations to be implemented on the basis of research contracts: 
funds necessary: 6 000 000 ECU. 
1.7. ATMOSPHERIC DISPERSION OF FISSION PRODUCTS AFTER AN ACCI­
DENT (Reference: point l.A.2.7 of the RAP) 
Objectives 
The problems associated with discharge of radioactive products from a nuclear power 
station in the event of an accident and with the atmospheric dispersion of those prod­
ucts are vitally important in the determination of the radiological consequences of such 
discharge both in the immediate vicinity of the power station and at regional level. 
These problems are particularly acute in Europe, where, owing to the size of the States 
and the proximity of frontiers to a considerable number of nuclear power stations, the 
consequences of discharges cannot be regarded solely within a national context. 
It is mainly for this reason that the Commission proposed research in this field under 
the first shared-cost research programme (1979-1983) (Area C of that programme). 
It should be stressed that the current studies in Area C of the first shared-cost pro­
gramme, just like those proposed for the second programme, are confined to aspects of 
the atmospheric transport of active pollutants and thus do not include the source term 
dealt with under heading 1.6 above or evaluation of the radiological consequences, for 
which other disciplines are required (')• Furthermore, these studies relate mainly to 
major hypothetical accidents which go beyond the design basis accident taken into 
account for evaluations of power station safety. For these accident categories, the zones 
concerned, the height above ground of the plume and the migration durations are 
appreciably more important than for a design basis accident, and a transition is made 
from the familiar field of meteorology/atmospheric dispersion on a 'local' scale (up to 
10 km) to that on at least a 'regional' scale (several tens of km). For this reason, and also 
because of the increased importance of partial effects (topography, inversions/stratifica­
tions, geostrophic entrainment, land-sea interface, etc.), the treatment of atmospheric 
dispersion no longer lends itself to the extreme simplifications of the local scale, and 
information, models and validations are also much less available in this case and are 
generally insufficient. Moreover, it should be noted that the available models are all 
deterministic. The 'variability', the semi-stochastic nature, of atmospheric transport is a 
very fundamental aspect which has so far been little studied. It is necessary to take vari­
ability into account mainly in the case of discharges which are short or of an intensity 
which varies with time, these being characteristic of accident situations. 
A small amount of work was undertaken under the first shared-cost programme. It 
seemed from the start that the resources and the time available would, at best, only ena­
ble the foundations to be laid for a longer-term coordinated project and that it would in 
any case be necessary to continue the work after 1984, a point of view which was also 
supported at the time by the ACPM (opinion of 27 March 1980). It was for this reason 
that the first programme was devoted, in particular, to cataloguing and more realistically 
evaluating the problems involved and the relative importance, at regional level, of a 
wide range of aspects and partial factors affecting the dispersion. 
It was thus that a certain amount of theoretical and experimental work concerning the 
effect of turbulence caused by buildings, the height of plumes above ground and the 
effect of the urban heat pocket, the on-shore breeze and dispersion by low winds was 
undertaken, but no major experiment 'in the field' was considered under the first pro­
gramme. Such projects will be proposed in the context of the second programme. 
(') Reference: Radiation-protection programme proposal. 
Official Journal of the European Communities 19.9.83 
The research work proposed below is in conformity with the recommendations of the 
ACPM study group responsible for monitoring the implementation of Area C of the first 
programme; when it is put in hand, account will be taken of the intention expressed by 
the various participants in the current programme to pool their resources for the purpose 
of carrying out certain sets of experiments both in the field and in wind tunnels. 
There is an obvious relationship between this research in the context of nuclear safety 
and the work carried out at Community level pursuant to Article 37 and for the pur­
poses of environmental and radiation protection. The existing links will be consolidated 
and the projected work will provide a valid contribution of broader interest in a field 
which is still insufficiently explored. 
Activities 
1.7.1. Discharge characteristics 
The parameters of an accidental discharge (duration, height above ground, etc.) some­
times condition the processes of subsequent atmospheric transport very significantly. 
This aspect, which apart from the modelling of hot jets, has not been touched upon in 
the current programme, will have to be dealt with in the second programme. 
1.7.2. Partial effects 
Studies in addition to those which are being conducted at present will be necessary. 
They will deal with dispersion above the sea and the transition land-sea(land), the influ­
ence of wind shear on lateral dispersion, dry or wet deposits, the influence of the dura­
tion of emission and transport, etc. These different studies will necessarily include 
experimental validations in the field on a limited scale and/or in wind tunnels or water 
tunnels (revolving). 
1.7.3. Dispersion modelling and validations 
Various models were prepared and sometimes validated in the first programme: on the 
basis of multi-annual statistical data. However, the importance of certain factors, topo­
graphy and shear in particular, will have to be better understood and this will require 
partial tests either in the field or in wind tunnels and water tunnels. 
Furthermore, in the current programme a study in a wind tunnel demonstrated the value 
of digital image analysis methods for studying the development and fine structure of 
simulated plumes under various weather conditions. Such studies should be continued, 
and should also be carried out at a theoretical and more fundamental level. It is likely 
that this work will result in the preparation of probabilistic models of atmospheric trans­
fer. 
1.7.4. Full-scale experimental validation 
Limited experiments were carried out under the first programme (tracers and meteorol­
ogical soundings) and mainly concerned partial effects or validation of models on a 
local scale, but no major 'in-the-field' experiment was performed. Such projects should 
now be considered; they will, of necessity, require the combined resources of several 
laboratories and will probably not be put in hand at the very beginning of the second 
programme. 
Certain aspects have already been mentioned in Sub-sections 1.7.2. and 1.7.3. above; 
diffusion above the sea, land-sea-(land) transition, influence of the local or regional 
relief. Other, at least partial validations of models on a regional scale (coastal sites and 
continental sites) are certainly necessary to supplement the various simulations consid­
ered, but will require detailed preliminary discussions with the study group concerned 
19. 9. 83 Official Journal of the European Communities No C 250/29 
and other bodies. Their extension to the intermediate scale (mesoscale, beyond 50-100 
km) would be conceivable at a later stage in association with other activities (radiation 
protection programme, etc.). 
Community contribution 
The studies or experimental validations that are planned will have to receive financial 
backing that takes account of the difficulties that beset experiments in the field and are 
caused by uncertainties in respect of weather and climatic conditions. 
Joint programme of investigations to be implemented on the basis of research contracts: 
funds necessary: 6 500 000 ECU. 
1.8. METHODS USED FOR PROBABILISTIC RISK EVALUATION (Reference: point 
l.A.2.1 of the RAP) 
Objectives 
In recent years, work in the field of risk analysis based on probabilistic methods has 
progressed considerably. 
Studies in this field were subsequently carried out in the United States in particular 
where the first major application of this analytical method was to the analysis of light-
water reactor safety (Wash 1400 or the Rasmussen Report). Further projects of a similar 
nature such as the German nuclear power station risk study (GRS 1980 or the Birkhofer 
Report) then followed. Work of the same type has since then been undertaken in other 
countries. Although quantitative risk assessment based on analysis of the probability of 
events still suffers from quite considerable uncertainties, this method forms the basis of 
the 'Proposed policy statement on safety goals for nuclear power plants' which the 
Nuclear Regulatory Commission published in February 1982. This probabilistic 
approach in safety analysis is of the greatest importance, since, for the first time, an 
attempt is being made to develop a method capable of assessing the level of risk asso­
ciated with a technical decision. 
Insofar as this method is successfully applied, its importance will go far beyond the 
nuclear field. It may be considered that, in the long term, it will be possible to apply it 
to the quantitative assessment and comparison of risks associated with various major 
activities in our industrial society. It will thus be able to contribute to creating a more 
objective definition of the acceptability of risks. In Europe, fruitful cooperation has 
been established between various institutes which had developed a probabilistic 
approach to risk assessment. This cooperation was facilitated by the role played by the 
JRC, which, in this field, possesses a team whose recognized skills are applied within 
the framework of the programme on reliability analysis, risk evaluation and data banks. 
However, where the methods used differ from one country to another, the problem of the 
acceptability itself of the probabilistic approach arises and will only be solved if the 
methods and models used are made more interchangeable. At the conclusion of a work­
shop held in Ispra in May 1982 on 'US PRA procedures, guide analysis and impact on 
European practices', the European experts present, with a view to coming closer to this 
objective, recommended that 'benchmark' validation and comparison exercises be car­
ried out with regard to specific aspects of probabilistic risk assessment. For example, an 
initial exercise, in which ten organizations are participating, is under way at present; the 
JRC is providing the technical secretariat and will draw the necessary conclusions and 
acquire the requisite information from it. 
This experiment must henceforth be more systematically performed. The shared-cost 
programme lends itself well to this type of 'benchmark' exercise, in which all the spe­
cialized teams that are operational within the Community should participate. The JRC, 
in close liaison with the managers of the shared-cost research programme, will continue 
No C 250/30 Official Journal of the European Communities 19.9.83 
to play a central role, both in organization and in the processing of the results. An initial 
choice of topics for the exercises under consideration will be defined in May 1983 at the 
meeting of the Group of Experts on Reliability (study group of the ACPM on Reactor 
Safety). 
Activities 
Between five and ten benchmark exercises (BE), which will be carried out at intervals 
during the programme, are proposed. The subjects are set out below in a general man­
ner. They will be defined in greater detail during the discussions with the groups of 
experts and participating teams. 
— Event tree: this BE is intended to analyze the various procedures by which system 
reliability evaluations are coupled together in the probabilistic analysis of an acci­
dent path. 
— Accident sequence analysis: the analysis mentioned above is also approached in 
this BE. Here the interaction of the accident development in terms of consequences 
and/or physical parameter behaviour and system reliability analysis will be 
accounted for. 
— Common cause/mode failure: this BE addresses the problem from one side of 
identification and classification of these failures, from the other side of their analyt­
ical treatment. 
— Reliability data: this BE will assess the procedures used to merge raw data from 
various sources in order to evaluate reliability parameters for components. 
— Human modelling: this BE approaches the important problem of human behaviour 
and data. 
— Probability of failure of structures and structures' response to seisms: these two BE 
are devoted to the analysis of methods able to examine the distribution of strength 
of structures versus load. 
— Consequence analysis: the aim of this BE is to compare various methods for the 
analysis of the transport of radioactive material to the environment. 
Community contribution 
Joint programme of investigation to be implemented on the basis of research contracts: 
funds necessary: 1 800 000 ECU. 
1.9. PARTICIPATION IN INTERNATIONAL RESEARCH PROJECTS OR IN PRO­
GRAMMES IMPLEMENTED OUTSIDE THE COMMUNITY 
The high cost of safety research programmes in which large-scale experimental facilities 
are used and the active international cooperation which has been established in this 
field inevitably bring about the setting up of international research projects other than 
those which have long been implemented and proposed within the framework of the 
Community. 
Three options are open at present: 
— the LOFT programme proposed by the USDOE for which a consortium was set up 
under the responsibility of OECD-NEA. This project extends the use of the LOFT 
19.9.83 Official Journal of the European Communities No C 250/31 
test reactor after completion of the programme sponsored by the NRC. A series of 
additional tests in the following fields has been planned: Large- and small-leak 
losses of primary coolant, secondary-side failures, release of fission products in the 
event of coolant loss, thermohydraulic studies concerning fuel deformation after 
loss of coolant. The consortium is open to the Community Member States and sev­
eral have already decided to participate in it. 
— the Marviken V Project proposed by Sweden, which concerns the behaviour in the 
primary circuit of a light-water reactor of volatile fission products and heavy aero­
sols generated during core meltdown. Many bodies in Community Member States 
and non-Community countries such as EPRI or Ontario-Hydro have announced 
their participation. 
— the 'Clean-up TMI-2' programme propsed by the USDOE, which covers a large 
number of subjects, including the following: the problems associated with the char­
acterization and transport of fission products in the containment, decontamination 
techniques, examination of the core and the internal parts. Several bodies in Mem­
ber States have decided to participate in this programme in various ways: second­
ment of staff, examination of specimens, etc. 
As can be seen from the foregoing, these three international projects cover tech­
nical subjects closely connected or directly related to the research proposals set out 
below for the second shared-cost research programme, particularly with regard to 
thermohydraulics and the severe damage suffered by the fuel in the event of a 
LOCA (LOFT, TMI-2) and to the source term resulting from fission products in 
accident situations (LOFT, TMI-2, Marviken). Furthermore, these three projects 
will be implemented during the 1983-1986 programme in parallel with the second 
shared-cost action programme (1984-1987). 
Effective participation by the Commission in all of these international projects or in 
specific parts of one or more of them has certain advantages: 
— It will enable the Member States that are not participating bilaterally in these pro­
jects to obtain full information on the results and the state of progress of the work. 
— Where certain specific points are concerned, it will enable the current shared-cost 
projects to benefit from results and information which will contribute to their suc­
cessful completion. 
— It will enable the results to be used in putting in hand 'benchmark' exercises for the 
validation of codes and models at Community level as part of the shared-cost pro­
gramme. 
Apart from the fact that the results of these international projects would be freely 
available to the Commission, effective Commission participation will make it pos­
sible to second the scientific staff of contractors participating in the shared-cost 
research programme to a given project, to participate in the analysis of the results, 
to examine specimens and to develop advanced measuring equipment in the labor­
atories of contractors of the shared-cost research programme. 
The Commission has examined various cases in which effective participation or cooper­
ation between the Commission and these projects would be possible within the frame­
work of the shared-cost programme propesed here. Two cases can be considered: 
— participation involving a financial contribution from the Commission, which 
would be limited and compatible with the funds allocated in the fields in question. 
In this case, the Commission could conclude contracts with the project sponsors 
after having consulted the 'CGC nuclear fission' (')• The appropriations necessary 
for the conclusion of these contracts would be acquired from the funds allocated to 
the cooresponding aspect of the shared-cost programme propsed here. This 
arrangement could be applied to obtaining full membership in the LOFT consor­
tium for the Commission. Preliminary contacts made with the USDOE indicate 
that Commission participation could be considered on the basis of an overall con­
tribution of one million ECU. 
(') Management and Consultative Committee set up in order to replace the ACMP Reactor 
Safety. 
No C 250/32 Official Journal of the European Communities 19. 9. 83 
— participation which varies in accordance with individual cases and is capable of 
being developed into full participation in the project, thus providing access to its 
results: such participation, however, would be negotiated on the basis of an 
exchange of technical information. The Commission would pay for its participation 
by making available to the sponsors of the projects the results and progress reports 
relating to certain parts of the present shared-cost programme which are of relev­
ance to the field dealt with in the project in question. Depending on the case, this 
provision of technical information could be accompanied by a financial contribu­
tion which, as in the first case, would have to be compatible with the sums allo­
cated to the technical field under consideration. The procedures involved in this 
type of arrangement would require that contracts be drawn up between the Com­
mission and the project sponsors, the conclusion of which would be subject to the 
opinion of the 'CGC nuclear fission'. This arrangement could be considered for 
Commission participation in all or part of the Marviken V Project. 
It should be noted that the solution based on making progress reports and the 
results of Community programmes available is not detrimental _to the interests of 
the Member States or, more specifically, of those who are already participating on a 
bilateral basis in the international projects in question. Indeed, the results and the 
state of progress of Community projects are described in communications pre­
sented at international meetings and at various conferences and symposia; further­
more, when the results of the Community programmes are officially published, they 
enter the public domaine. The procedures proposed above only accelerate the pro­
cess of dissemination of the Community research results, and this can only improve 
the exchanges of technical information between Member States and non-Com­
munity countries. 
Commission participation in the 'Clean-up TMI-2' programme could be considered 
on the basis of either one of the two possible arrangements described above; how­
ever it should be noted that provision has already been made for the participation 
of the JRC laboratories in the analysis of certain specimens. 
The two arrangements set out above for the purpose of obtaining Commission participa­
tion in international research projects can also be used in negotiating access to the 
results of research work conducted at national level in certain non-Community coun­
tries. Consideration might be given, for example, to the EPRI and Sandia programmes 
on hydrogen, the Sandia programme on the source term and the ROSA IV programme 
on thermohydraulics, which are closely related to points 1.5, 1.6 and 1.4 above and, in 
particular, to the programmes on severe fuel damage like PBF, EPRI (USA) NRU (Can­
ada) already mentioned under points 1.4 and 1.4.4 above. 
SECOND PART 
2. LIQUID METAL FAST BREEDER REACTOR (LMFBR) SAFETY 
COST SHARING PROPOSAL 
INTRODUCTION 
Most of the Member States of the Community have devoted substantial efforts to the develop­
ment of Liquid Metal Fast Breeder Reactors during the last twenty-five years. During this period 
a very large financial effort has been devoted to this development. Today fast breeder develop­
ment expenditure still accounts for approximatety 20 % of total R, D + D expenditure in the 
energy sector. Against this efforts one must record impressive technical achievements. Not only 
several experimental reactors and prototype reactors have ben succesfully constructed and oper­
ated, but also one large station (1 200 MW(e)) is in construction and nearing completion. These 
achievements are so far unmatched in any other regions of the world. 
19. 9. 83 Official Journal of the European Communities No C 250/33 
The following table summarizes the reactor projects achieved or under consideration according to 
countries or groups of collaborating countries. The date shown on brackets is the date of begin­
ning of operation. 
Country 
United Kingdom 
France 
Federal Republic 
of Germany 
Italy Experimental and 
test reactors 
DFR(1963) 
Rapsodie(1967) 
KNK II (4) 
(1977) 
PEC (1986) Prototype 
(200-300 MW) 
PFR(1974) 
Phenix(1974) 
SNR 300 (2) 
(1986) Demonstration plants 
(1 200MW(e)) 
CFR project not yet 
adopted 
Super-Phenix (') 
(1984) 
SNR 2 (3) project 
not yet adopted 
(') In collaboration with Italy, the Federal Republic of Germany, Belgium and the Netherlands. 
(2) In collaboration with Belgium and the Netherlands. 
(3) In collaboration with France, Italy, Belgium and the Netherlands. 
(4) In collaboration with Belgium and the Netherlands. 
The table shows that a trend has developed towards cooperation amongst several Member States 
for the construction of demonstration plants. Important cooperation agreements have been con­
cluded amongst research organizations and industrial partners (electricity producers and design 
and construction companies) from several Member States. 
Outside the Community, most of the industrialized countries of the world have also devoted large 
efforts to fast breeder reactors. 
In the USA the 200 MW(th) experimental EFFBR was operated from 1963 to 1972 and provided 
valuable experience with various reactor systems. The 62,5 MW(th) experimental reactor EBR-II 
has been in operation since 1965. Construction of the FFTF experimental reactor (400 MW(th)) 
which, due to its power will practically function as a demonstration plant, has been completed in 
1980, Development and experimental studies on equipment development, sodium technology, 
fuel cycle and confirmation of licence requirements have been carried out. Construction of the 
Clinch River Breeder Reactor (CRBR) was suspended during the Carter Administration, and the 
prospects of this demonstration project are still uncertain today. 
In Japan the experimental fast reactor 'JOYO' with a projected power of 100 MW(th) (Spring 
1983) is in operation since 1977. Design of the 300 MW(e) prototype reactor 'MONJU' has been 
completed and its construction has been recently decided. Preliminary designs of a larger demon­
stration reactor are being made. The construction of such a demonstration reactor is envisaged to 
start after one year of operation of 'MONJU' and to be followed by serial construction of several 
near-commercial reactors which will be similar both in size and design to those of the demonstra­
tion reactor. 
In the USSR development and construction of fast reactors is a constituent of the power econ­
omy development of the country. At present, two experimental reactors, the BOR-60 (60 MW(th)) 
and the BR-10 (10 MW(th)) are operating. The demonstration reactor BN-350 has been success­
fully operating at 350 MW(e) of it being expended in desalting seawater since 1973. The second 
demonstration reactor BN-600(MW(e) is operational since 1981. The design of the BN-1600 
commercial fast reactor (1600 MW(e)) is at the development stage. The possibility of increasing 
the power of the BN-600 type reactor to 800 MW(e) is under consideration. 
COMMUNITY RESEARCH ON FAST BREEDER REACTOR SAFETY 
For fast reactors to become acceptable, their performance in terms of safety, radiological protec­
tion and impact on the environment in normal and accidental conditions must be shown to be 
equivalent to that of the then established thermal reactors. 
No C 250/34 Official Journal of the European Communities 19. 9. 83 
It is one of the primary tasks of the fast reactor demonstration and safety programmes now in 
hand in various countries of the Community to show that this condition can be satisfied. 
The experience accumulated during the exploitation of prototype fast reactors (Phenix, PFR) 
shows that it is possible for a fast reactor to operate within the general siting constraints relating 
to controlled radioactivity discharge normally applied to thermal reactors. Doses experienced by 
operators have also been satisfactorily low. 
However, the exploration in depth of the various possible safety issues which could affect future 
fast reactors, namely: 
— the prevention of the growth of small incidents into accidents; 
— the identification of accident initiation and the description of the subsequent transients and 
their effects on key plant components; 
— the radiological consequences of accidents, i.e. the internal redistribution of radioactivity 
following an accident and the definition of source terms for an external risk assessment, 
requires continuing and diverse efforts in software development and in production of reliable 
physical data for design and performance. 
Safety programmes with the objective outlined above are a vital complement to the construction 
and operation of prototype and demonstration reactors. 
The Community has so far supported fast breeder reactor safety in two ways: 
(a) by the execution of the Community's own research programme in the Joint Research Center 
(JRC) 
(b) by activities aimed at improving coordination and collaboration amongst national pro­
grammes as well as with the JRC programme through the Fast Reactor Coordinating Com­
mittee (') 
Concerning (a), the current activities of the JRC, i.e. those performed under the 1980-1983 plu-
riannuel programme, are subdivided in three main chapters, i.e.: 
Accident initiation and transition phase 
The activities are concentrated on liquid metal boiling theoretical and experimental studies, on 
the development of the European Accident Code and on fuel/coolant interaction research. The 
objective of the liquid metal boiling studies is to gain data on coolant behaviour in case of anom­
alous operating conditions such as blockage in a rod bundle, flow run down due to pump failure 
or power excursions. The European Accident Code is a modular system of computer codes des­
cribing the different phases of hypothetical accidents; the pilot version of this code is already 
operational. The subjects of the fuel/coolant interaction research are the development of physical 
models and codes, experimental studies of factors influencing the interaction process or the trig­
gering of vapor explosions and the verification of theoretically postulated mechanisms of vapor 
explosions. 
(') Committee created by the Council in April 1970 with the mandate 'to work out and imple­
ment plans for coordination and cooperation on the broadest possible scale between the var­
ious programmes by means of the most suitable procedures and to make any helpful sugges­
tion in this connection'. 
19.9.83 Official Journal of the European Communities No C 250/35 
Accident post-disassembly phase 
The activities in this area are devoted to a more realistic description of the post-disassembly 
phase, to the analysis of the behaviour of the primary containment system and to post-accident 
heat removal (PAHR). A start has been made to evaluate the capabilities and the models of the 
US code Simmer 2 in the context of post-disassembly calculations and to investigate the possibil­
ity of an experimental programme for its validation. 
The code validation programme for containment (COVA) has been virtually completed. To 
obtain better agreement between experiments and numerical results, finite element codes have 
been imposed. The code validation for subassemblies (COVAS), aimed at validation of structural 
codes for dynamic plastic analysis is continuing. 
Post accident heat removal studies are supported by both out-of-pile and in-pile experiments. 
Out-of-pile experiments, to be performed in the FARO facility now being completed, aim at stu­
dying fuel/coolant interaction and PAHR phenomena under realistic accident conditions using 
real reactor materials. The in-pile experiments (in the US, Sandia, at Grenoble, Melusine reactor 
and at Mol, BR 2 reactor) aim at demonstrating the coolability of core debris which may form in 
a severe hypothetical accident and settle on various parts of the reactor vessel. A comprehensive 
programme to develop and verify physical models and codes used to predict the temperature 
field in a debris bed and in debris retention devices is also performed. 
Material research 
Research in particular on stainless' steel is performed at the JRC in several areas: fracture 
mechanics with particular emphasis on irradiated materials; study of creep crack growth for aus-
tenitic steel AISI 304 and 316 for typical operating conditions (load, temperature, creep-ductility 
etc.), studies on material dynamic behaviour and definition of corresponding constitutive laws. 
The latter studies were fundamental for the Cova and Covas programme. Finally they should 
provide an understanding of the response of real reactor structures under different loading condi­
tions (temperature, stress state) and various degrees of degradation (welding, creep, mechanical 
fatigue, irradiation). A large high load dynamic test facility is under construction at Ispra. It will 
be used to investigate how the results obtained for small specimens (up to 20 mm2 cross section) 
can be transferred to large structures of damaged materials (up to 5 000 mm2). 
The activities of coordination performed at the Brussels Headquarters are mainly focussed on the 
following fields of safety: 
— With the assistance of the Safety Working Group (SWG), which is an expert group of the 
Fast Reactor Coordinating Committee, progress has been maintained in the preparation of 
common safety criteria and guidelines for fast reactors. 
The Whole Core Accident (WAC) sub-group of the SWG, which also acts as expert group in 
the frame of the Advisory Committee on Programme Management on Safety originated the 
European Accident Code and has continued to advice on its progress; it has also promoted 
comparative calculations of European as well as US codes for selected core accidents (tran­
sient over-power and loss of flow). The Containment Loading and Response (CONT) sub­
group of the SWG, which also acts as an expert group in the frame of the Advisory Com­
mittee on Programme Management on Safety has dealt with the questions of behaviour of 
primary containment and internal structures during a hypothetical core disruptive accident; 
a critical assessment of the adequacy of mathematical tools available has been performed. 
This sub-group also followed closely on-going work to assess the consequences of sub­
assembly accidents on to the adjacent structures, and has more recently considered secon­
dary containment problems with a view to a realistic assessment of radioactivity source term. 
No C 250/36 Official Journal of the European Communities 19.9.83 
— With the assistance of the Codes and Standards Working Group (WGCS) which is another 
expert group of the Fast Reactor Coordinating Committee, steady progress is being main­
tained in the progressive assessment of divergencies existing between the various codes and 
standards applied in the Community to the design, construction and quality control of fast 
reactor components as well as in materials specifications. The Group has been systematically 
comparing and evaluating rules, design codes, material specifications and materials data 
from Member States and, when possible, from third countries. It has also conducted a survey 
of the status of non-destructive methods for in-service inspection in LMFBRs. 
DEFINITON AND PREPARATION OF THE COST SHARING PROGRAMME 
The Commission proposes that shared cost action should also be employed in future as a means 
to reinforce coordination of national programmes and to complement the already existing action 
oftheJRC. 
The preparation of the LMFBR cost-sharing research programme for 1984-1987 started towards 
the end of 1982 with the assistance of the Safety Working Group of the Fast Reactor Coordinat­
ing Committee. The broad aim in the selection of topics has been that the programme should be 
consistent with the objectives of the Council Resolution of 18 February 1980 concerning fast 
breeder reactors (') and in particular should contribute to complete and to add value to the 
researches undertaken in the member countries and, if such should be the case, fill the existing 
gaps. 
The abovementioned Council Resolution underlines the importance of the fast breeder option for 
the future energy supply of the Community, stresses the importance of continuity in the effort of 
developing and demonstrating the system, reaffirms the paramount importance of safety as an 
objective of the development and demonstration effort and calls upon the Community to lend 
support for the above objectives. 
Additional criteria for selection of topics have been: 
— to maximize effectiveness of Community support by selecting a limited number of important 
areas where Community support can best stimulate coordination and collaboration amongst 
national programmes, 
— searching for common interest of Member States, avoiding design-specific issues, 
— giving adequate emphasis to the areas of accident prevention, without however neglecting 
the areas of accident analysis and mitigation (including severe accidents), 
— making sure that no duplication of effort occurs in those topics which are addressed by the 
JRC, and that direct action and cost-shared action complement and enhance each other. 
The proposed programme is articulated in seven programme units as follows: 
1. Instrumentation, control and protection; 
2. Transient analysis (operational safety); 
3. Integrity of components and structures; 
4. Safety aspects of sodium technology; 
5. Fuel behaviour and post-failure phenomena (in-pile experiments); 
6. Fission product transport in severe accidents; 
7. Molten material motion and interaction in severe accidents. 
(•) OJ No C 51, 29. 2. 1980, p. 5. 
19.9.83 Official Journal of the European Communities No C 250/37 
Programme units Nos 1, 2, and 3 are essentially concerned with accident prevention. The three 
units of this group emphasize: 
— the early recognition of faults and the prevention of their evolution into accidents, 
— improved description of operational thermohydraulic transients with a view of a better defi­
nition of operating margins, 
— The reliable design of structures important for safety with a view of an improved prediction 
of component's lifetime and safety margins. 
Programme units Nos 4 and 5 are concerned with work tending to resolve particular issues in the 
analysis of those accidents which, though potentially severe, do not fall necessarily into the cate­
gory of whole core accidents. Thus their objectives are: 
— to examine the safety aspects of sodium as a coolant with a view to both exploiting to the 
full its characteristics as a good coolant by improving the modelling of flow, particularly 
under fault conditions (natural convection) and to improve the assessment of the consequ­
ences associated with the handling of large quantities of sodium coming accidentally into 
contact with air, water and concrete. 
— to improve knowledge of transient fuel behaviour and fuel failure modes. Also to improve 
knowledge of in core post-failure phenomena, particularly with a view to verify the condi­
tion under which a sub-assembly accident could propagate to adjacent sub-assemblies. 
Programme units Nos 6 and 7 are concerned with the description of the consequences of a severe 
core damage. Their objectives are: 
— the description of radioactive material distribution within the containment following severe 
accidents, the evaluation of containment and the generation of the source term required by 
codes used to calculate any ensuing offsite damage; 
— the description of molten core material motion and interaction following severe damage. 
INSTRUMENTATION CONTROL AND PROTECTION (Reference: point l.B.2.1 
of the RAP) 
Objectives 
The primary goal of this programme unit is to enhance safety by preventing accidents. 
This will be achieved by reducing the frequency and the severity of the demands being 
made upon reactor structures and to improve reactors' ability to respond and recover 
from load following and other transients without exceeding safety and material damage 
limits. 
Designers of LMFBR control and protection systems are faced with space and complex­
ity problems: space because the core is small in relation to the number of instruments 
required to meet both safety and availability targets in the LMFBR environment: com­
plexity because combinations of measurements of different variables are desirable to 
provide the best balance between the conflicting risks of failure to trip and unnecessary 
action. 
The availability of small, relatively simple, low-cost programmable logic elements offers 
the opportunity to achieve enormously enhanced flexible signal handling, processing 
and correlating capacity; high reliability and high availability at a reduced cost, while at 
the same time overcoming some of the problems of component obsolescence e.g. availa­
bility of replacement parts. 
No C 250/38 Official Journal of the European Communities 19. 9. 83 
However, if the maximum benefit is to be obtained from micro-electronics it is not suffi­
cient to merely replace hard-wired logic elements by micro-computers; one's whole 
approach to control and protection system design and licencing needs to be re-exam­
ined. 
Secondly, computer-based systems have their own particular problems, the most impor­
tant of which is widely recognized to be the specification and reliability of software; 
others, are materials unfamiliar in nuclear applications and novel signal transmission 
methods. 
The opportunities offered were surveyed for the Commission and described to the 
LMFBR Safety Working Group in 1980. Since then four studies have been implemented 
to confirm interest and opportunities, generate programme topics and establish links 
between experts in electronic and nuclear industries and other users of highly reliability 
computers. A fifth study showed that applications for the techniques of artificial intel­
ligence can be anticipated in the design of decision making and diagnostic systems and 
the modelling of operator behaviour. The latter two items are now part of the LWR saf­
ety programme activity entitled, 'Human factors' and 'Man-machine interaction', which 
has been structured to complement the activities described below. 
The programme is divided into tasks covering sensor improvement, signal processing 
and system design. Emphasis will be on validating and extending applications of com­
puter science in nuclear safety. 
Sensor improvement 
Development of improved sensors for measurement of temperature level, flow, impurity 
level, neutron flux, radiation, sound, displacement, etc. 
The specific improvements sought are: 
— for thermocouple and flow sensors, longer life to reduce the number of spares 
required to guarantee adequate redundancy between access periods and positive 
failure indication to indicate actual redundancy; for instrumentation in the core 
vessel, easier handling a reduction in the number of electrical leads passing through 
the reactor roof and the area between the top of the core and the roof, and 
improved discrimination against environmental interference; 
— for sensors used for accident monitoring, recovery guidance and damage assess­
ment, broad range, resistance to hostile environment and agreed qualification 
methods. 
Computer based signal processing techniques and applications 
Development of signal processing techniques for surveillance, inspection, control and 
protection by the application of computer science and technology developed outside 
the nuclear industry to LMFBR problems, and the validation of the application in a 
LMFBR context. Results of a quality that would permit a high degree of automation in 
the use of these signals and hence relieve the operator of the need to examine large vol­
umes of data are sought. 
System design 
Development of system design concepts that reflect general trends toward distributed 
control, more elaborate control structures and increasing levels of monitoring and meet 
requirements for demonstrable reliability, safety, speed of operation and improved plant 
availability. 
19.9.83 Official Journal of the European Communities No C 250/39 
Activities 
2.1.1. Sensor improvement 
For core instrumentation four approaches aimed at enhanced safety, reliability and eas­
ier handling and location are considered appropriate: 
— development of sensors using novel or diverse measuring principles: novel flow, 
temperature and pressure sensors; 
— the further improvement of discrimination against radiation in neutron flux sen­
sors, improvements in the thin sheaths separating thermocouples, strain gauges or 
acoustic transducers from sodium; 
— grouping of instruments into a single easy-to-handle probes; 
— exploring the use of magnetic, ultrasonic or other wireless means for transmission 
of information from a primary sensor, which may be situated on a sub-assembly or 
in an inaccessible place to a more conveniently placed receiver. 
2.1.2. Computer-based signal processing techniques and applications 
The aim is to obtain more information from existing sensors. In general emphasis will 
be laid upon the applications of techniques such as pattern recognition and adaptive 
control to demonstrate their value to LMFBR: 
— thermal and acoustic noise analysis e.g. from boiling or loose parts; 
— improvement of discrimination between faults and artifacts in acoustic and eddy-
current inspection; 
— automation of failed fuel location using delayed neutron detectors and other avail­
able information; 
— reactivity balance and sub-assembly state monitoring. 
The programme covers research, development and demonstration to prototype stage. 
Account will be taken of experience already acquired in existing experimental facilities. 
2.1.3. System design 
— Local area networks. 
The use of local area networks offers much reduced cabling and space demands, 
high capacity and the ability to select signals for correlation with a great deal of 
flexibility. 
The following stepwise approach leading to, but not including demonstration of a 
trial system is proposed: 
— investigation of trade-offs between network reliability; fault tolerance based 
upon redundancy and fault diagnosis and automatic reconfiguration; fail safe 
systems; access delays and message delivery time: maintainability, space and 
cost; 
— evaluation of what signals should be put on the network in relation to the 
extent of prior processing, the need for cross correlation; 
— assessment of environmental limitations on the use of optical fibres; 
No C 250/40 Official Journal of the European Communities 19. 9. 83 
— construction of an experimental three node network for performance assess­
ment using existing communication test facilities. 
— integrated microprocessor based surveillance and control system. 
Detailed specification and construction of an experimental integrated core surveillance 
system will be preceeded by a series of comparison and optimization tests to assess the 
relative performance of the various techniques being proposed for use of an integrated 
system and to justify their inclusion. An agreed set of test data simulating a developing 
sub-assembly fault will need to be generated. 
The experimental integrated core surveillance system will use results from all other 
activities and the work on information communication described above. It will include 
output generating and display system appropriate to a fast reactor. Simulated test sig­
nals and recorded data from real reactors will be used to test the system. It may be pos­
sible to arrange to feed the system with signals from a reactor. 
Community contribution: 
3 000 000 ECU. 
2.2. PLANT TRANSIENT ANALYSIS (OPERATIONAL SAFETY) (Reference: point 
l.B.2.3oftheRAP) ' 
Objectives 
The accurate knowledge of the parameters involved in transient phenomena and their 
variations during the transient is an important prerequisite both for studying the poten­
tial of the transient to degenerate into an incident and to evaluate the effect of the tran­
sient on the affected structures: for example an accurate assessment of temperature 
variation in a particular part of the plant serves to provide the correct input for the 
assessment of the effects of this parameter onto the structural integrity of the affected 
components. 
Within the limited resources assigned to this project it is possible to investigate only a 
limited part of the subject, which is potentially very wide in view of the complexity of 
the plant and the feedback that each part of the plant, including the control system, pro­
vides to the others. 
Thus the main objective is to examine the transient thermohydraulics of few key compo­
nents, such as large vessels, pipes and heat exchanging equipment. For these, tempera­
ture and velocity distribution during operational transients (for example, start up, shut 
down and load variations) are at present calculated with assumptions which tend to 
affect these structures with unnecessarily high thermal loads. Therefore it is necessary to 
improve the existing codes and to validate them with experimental results. 
Improvements to existing modular codes capable of simulating either pool or loop 
sodium cooled reactor plant behaviour following thermohydraulic and power disturb­
ances will also be started, especially for those modules corresponding to the key compo­
nents mentioned above. 
Activities 
Thermohydraulic in large vessels, pipes and heat exchanging equipment: 
19. 9. 83 Official Journal of the European Communities No C 250/41 
— improvement and development of existing codes for mixing and stratification in 
large plena, pipes and headers, and validation; 
— insertion of plena models into plant dynamic codes; 
— theoretical and experimental investigations with regard to thermal striping prob­
lems; 
— improvement to existing modular codes for transient analysis. 
Community contribution: 
1 700 000 ECU. 
2.3. INTEGRITY OF COMPONENTS AND STRUCTURES (Reference: point l.B.2.2 
of the RAP) 
Objectives 
Structural integrity under both normal and fault conditions, is an important element of 
reactor safety. Structural integrity must be assured for adequately long periods of time 
under normal conditions and a clearly defined set of fault conditions. 
For liquid metal fast breeder reactors, the components and structures considered in this 
proposal are those which are or can be in contact with the liquid metal, its vapour or 
blanket gases. 
A combination of a low operating pressure and high coolant thermal conductivity 
results in fast reactor structures and components which are often thin walled and rela­
tively flexible; furthermore the use of austenitic steel is much more frequent than in 
light-water reactors and operating temperatures are also higher than in light-water reac­
tors. Thus design and calculation codes as well as non-destructive methods of inspec­
tion must take into account phenomena and material properties for which experience in 
light-water reactor components design, construction and operation is not necessarily 
sufficient or relevant. 
Though a great deal of research effort has been devoted to material and structural 
aspects of LMFBRs, a number of areas remain in need of attention. 
The working group codes and standards of the Fast Reactor Coordinating Committee 
has been systematically comparing and evaluating rules, design codes, computer codes, 
material specifications and materials data from Member States and when possible from 
third countries. It has also conducted a survey of the present development status of non 
destructive testing methods for in-service inspection in liquid metal fast breeder reactors 
within the Community. The working group has made sufficient progress to permit the 
identification of experimental programmes, which could usefully be executed by cost 
sharing actions. 
The Expert Group CO NT (containment, loading and response) of the Fast Reactor Saf­
ety Working Group has examined on-going activities in the field of dynamic material 
behaviour and structural codes in connection with severe accidents. 
It is to be emphasized that all the activities proposed in this programme unit acquire 
tests on large specimens and/or model structures which can only be performed in 
specialized large test equipment available in the Member States. 
The proposed research represents an important extension of the direct research pro­
gramme funded by the Commission at the JRC, dealing in particular with creep pheno-
No C 250/42 Official Journal of the European Communities 19.9.83 
mena, creep-fatique interaction, fracture mechanics, dynamic material behaviour and 
structural analysis. It is foreseen that the JRC specialists will be strongly involved in the 
collaboration with the national laboratories participating in this programme. 
The objectives and justification of the activities proposed for the 1984-1987 cost sharing 
programme follow. 
Material properties, structural analysis and validation of codes 
— Constructive modelling in the range of inelastic deformations. 
The design of fast reactors requires a great number of inelastic analyses in order to 
predict the behaviour of structures which have complex shape and operate at high 
temperature and under important cyclic loading. 
Fundamentals of inelastic analysis method exist in mathematical descriptions of 
material behaviours (constitutive equations) and their application procedure. Con­
stitutive equations, although much investigated, are not completely verified from 
the view point of design practice, due to substantial complexity of material behav­
iour itself at elevated temperature. 
— Service life prediction. 
The behaviour of components at high temperatures is significantly influenced by 
the operational history and by the initial stress distribution. The design is carried 
out with expensive calculation models and the results obtained depend strongly on 
the materials data used. It is very important to gain precise knowledge on how the 
component's remaining life can be predicted by an appropriate intermediate mater­
ial surveillance test programme and calculations. This will help in the correct 
assessment of the safety margin of components during their service. The relevance 
of results from such tests on specimens or components (structure, mechanical pro­
perties, crack, strain, non-destructive testing) within the framework of a severe life 
surveillance programme has been scarcely verified and interrelation between such 
results and the calculation procedures needs to be established and validated. This is 
especially true for weldments. 
— Dynamic analysis (only for this paragraph ref. point l.B.2.5) 
In the case of internal or external events such as steam generator accidents, core 
accidents, seismic events, etc , the reactor components and particularly the reac­
tor vessel and piping systems and supporting structures can be subjected to severe 
mechanical loadings. Advanced computational tools are already available, but fur­
ther effort is needed to improve the description of specific aspects. 
Structural defects propagation 
Crack propagation at high temperature has received comparatively little attention, with 
respect to the large amount of work which has been undertaken on structure phenomena 
below the creep range. 
For LMFBR components which operate at high temperature, the growth of cracks 
through a creep mechanism, or, where the component is subject to severe cyclic loading, 
through combined creep and fatigue mechanism, is an important phenomenon to be 
taken into account to achieve safe design. 
Although it is important to establish crack propagation data for the material used in the 
component by laboratory tests, this is in itself far from being sufficient. Assessment of 
component integrity requires that these data shall be applied to a structure of more com­
plex shape, where stress system differs significantly from that in the laboratory tests. 
19.9.83 Official Journal of the European Communities No C 250/43 
Activities 
2.3.1. Material properties, structural analysis and validation of codes 
— Constitutive modelling in the range of inelastic deformations: 
— improvement of the constitutive equations developed so far and application to 
the analysis of reactor structures taking account of the essential materials' pro­
perties and their evolution with time, temperature, loadings and irradiation; 
— validation of the method of analysis thus developed by experimental tests for 
shapes, materials (including weldments) using loading combinations which are 
representative of real operating conditions of fast reactors. 
This activity will be performed in collaboration with the JRC where an effort on the 
development of constitutive equations (in particular for damaged materials condi­
tions) is under way. 
— Service life prediction 
It is proposed to take as reference cases, for instance, one component for the pri­
mary or secondary circuit in stainless steel material (for example a highly stressed 
pipe elbow) and one component in ferritic material for use in the boilers (for exam­
ple tubing sections). 
The following programmes would be executed for both cases: 
— creep testing of base materials and weld materials after long term thermal 
exposure, fatigue testing at operating temperatures; 
these tests are to be executed both for a component of representative geometry 
and size and for the related specimen of a typical service life surveillance pro­
gramme ; 
— determination of residual strength properties after fatigue, creep of thermal 
exposure; 
— service life assessment of component by comparing theoretical and experimen­
tal values deducted from the results of the above experiments. 
— Dynamic analysis (only for this paragraph ref. point 1 .B.2.5) 
— development and improvement of codes for the design of large, thin walled 
vessels and piping systems against dynamic loadings (particularly in view of 
the fluid/structures interaction and buckling problems). 
— validation of the above codes with experiments: particular in order to deter­
mine the consequences of dynamic events to representative structures and the 
limits of their load-carrying capacity. 
The development of codes to assess the structural behaviour under dynamic loading, 
also in the case where fluid-structure interaction must be taken into account, has 
reached an advanced stage during the last years. However, the improvement of modell­
ing of specific aspects, e.g. roof impact during HCDA, modelization of in-vessel struc­
tures, description of buckling phenomena, need further development of codes and vali­
dation experiments on scale models. 
The JRC by means of the Cova programme and in the frame of several international 
collaborations, developed relevant competences in this domain. As for other areas, the 
JRC will have a central role in the research orientations to assure the complementarity 
of the different efforts. 
2.3.2. Structural defects propagation 
— Influence of residual and thermal stresses in the formation and propagation of 
fatigue cracks in elastoplastic conditions: 
No C 250/44 Official Journal of the European Communities 19. 9. 83 
— study of the initiating causes of cracks (such as welding cladding, etc.) and 
their classification in order of importance; 
— study and experimental tests on fatigue resistance of the component or struc­
ture. 
— Fatigue crack growth: 
— study and experimental tests on crack propagation through the walls of austen-
itic pipes under realistic loading conditions. Assessment of leakages through 
the cracks and their relevance to the structural integrity of the pipes (leak-
before-break criterion). 
— Tests on large specimen and model structures representative of large welded reactor 
structures, subjected to typical reactor loading histories, such as: 
— biaxial loading, combined membrane and biaxial loading, combined primary 
(mechanical) and secondary (thermal, residual) stressing, cyclic loading. 
Community contribution: 
7 700 000 ECU. 
2.4. SAFETY ASPECTS OF SODIUM TECHNOLOGY (Reference points l.B.2.3 and 
l.B.2.6oftheRAP) 
Objectives 
The objectives of this programme unit is to examine safety aspects of sodium as a coo­
lant with a view to both exploiting to the full its characteristics as a good coolant by 
improving the modelling of flow, particularly under fault conditions and to improve the 
assessment of the consequences associated with the handling of large quantities of 
sodium, which could come accidentally into contact with air, water and concrete. 
Sodium coolant flow modelling (point l.B.2.3) 
Experiments performed in prototype LMFBRs between 1978 and 1980 focussed atten­
tion! on the removal of heat by natural convection and led to international discussion of 
the limits to which it could be exploited. Still further extension of this very important 
safety margin is believed to be possible by improving the modelling of flow: specifically 
in large plena and a blocked fuel subassembly. Boiling in a sub-assembly can be an 
important aid to natural circulation initiation. Experimental verification of new flow 
models will be required. 
The result sought is a valid code that can be used in the safe design of LMFBR cooling 
systems. 
The Joint Research Centre is expected to continue its work on boiling in sub-assemblies 
under conditions representative of full power and possibly decay heat removal condi­
tions. 
Reactions of sodium with air, water and concrete (point 1 .B.2.6) 
Assessment of the risks associated with the large scale handling of sodium requires the 
consequences of sodium coming accidentally into contact with air, water and concrete 
to be understood under a wide range of conditions. 
19.9.83 Official Journal of the European Communities No C 250/45 
Though sodium fires have been investigated for some time a generalized code able to 
describe spray and pool fires in vessels of different sizes without excessive conservatism 
is lacking. The experimental programme prepared has as its objective a generalized des­
cription of the damage potential of a fire caused by sodium entering an air-filled space 
as a spray or a jet. Parameters to be taken into account are ambient temperature, pres­
sure and composition data, vessel geometry, the degree of dispersion of the sodium jet 
or spray. 
The particular feature of sodium water reactions to be studied is what happens when 
sodium enters a pool of water in a closed air-filled room, such as the room surrounding 
the steam raising plant of an LMFBR. Information is needed about hydrogen-oxygen 
and explosion limits and effects in gas mixtures heavily loaded with sodium oxide (or 
hydroxide) particles; the effect of the pressure generated by the reaction products on the 
sodium-water reaction rate and the hydrogen rate and the effect of varying overall 
sodium/water ratios, pool geometry, the shape of the reaction zone and the mixing 
model. 
The desired result is a code able to describe this very complex scenario well enough to 
evaluate the risk of structural damage and sodium release. 
The aim of the activity relating to sodium concrete reactions is to be able to predict the 
penetration of structures. Bare concrete and concrete covered with a defective steel liner 
will be studied experimentally and modelled as function of the water content of the 
concrete, sodium liner and concrete temperatures and pressure. 
The work described under this heading relates only to sodium and not to sodium-fuel 
mixtures. The description of the behaviour of the aerosol formed by a sodium fire is 
treated in the programme unit dealing with fission product transport because of its 
importance as a flocculant for radioactivity. 
Activities 
2.4.1. Sodium coolant flow modelling (point 1 .B.2.3) 
— Improvements to the modelling of natural convection and sodium boiling under 
shut-down conditions. Special attention will be paid to description of the heat path 
between the core and the intermediate heat exchanger, the temperature field in the 
upper plenum and the performance of the external heat dump and the intermediate 
heat exchanger. 
— Validation of the models described above. 
— Validation of codes describing the cooling of normal, demaged and blocked sub­
assemblies under the following condition: 
— low forced flows; 
— natural convection. 
at shut-down conditions. 
This activity will be expected under strict coordination with the JRC work on 
sodium boiling. 
2.4.2. Reaction of sodium with air, water and conrete (point 1 .B.2.6) 
— Experimental study and modelling of sodium fires initiated by sodium sprays or 
jets. 
No C 250/46 Official Journal of the European Communities 19.9.83 
— Experimental study of sodium-water-air reactions in a closed room, to include a 
study of hydrogen-oxygen explosion limits in the presence of a dense mixture of 
steam and 'sodium' aerosol. 
— Development of a global one dimensional model. 
— Experimental study and modelling of sodium reaction with bare concrete and con­
crete covered with a defective steel liner. 
Community contribution 
3 800 000 ECU. 
2.5. IN-PILE EXPERIMENTS TO INVESTIGATE TRANSIENT FUEL BEHAV­
IOUR AND POST FAILURE PHENOMENA (Reference: points l.B.2.3, l.B.2.4 of 
the RAP) 
Objectives 
Transient fuel behaviour 
A knowledge of transient fuel behaviour is essential for the determination of fuel failure 
and the subsequent accident evolution. While a number of codes exist for the detailed 
description of the fuel behaviour under normal operating conditions, codes treating the 
transient fuel behaviour have not yet reached the same level of sophistication. This is 
essentially due to the fact that for the analysis of hypothetical core disruptive accidents, 
on which for a long time the main effort was concentrated, the prefailure fuel behaviour 
is of lesser importance. As now increasing emphasis is given to a more realistic accident 
assessment prefailure phenomena get a higher priority. 
Transient fuel behaviour codes developed so far have generally been validated using 
results from experiments performed in pulsed reactors like Cabri or Treat. Most of these 
experiments simulate rather fast transients corresponding to loss of flow or overpower 
type accidents leading to fuel pin failure. Additional experimental data are necessary 
for slower transients. It is proposed to perform slow overpower transients in the HFR 
reactor at Petten in which similar experiments had already been carried out for loss of 
flow conditions. HFR is particularly suited for single pin transients with power dou­
bling times of one to 10 sec. The reactor has the possibility to precondition fuel pins 
over a certain period at nominal power before the transient is initiated. 
The loop necessary for the experiments is already existing. ECN Petten also posesses 
facilities needed to perform post-irradiation examination which is necessary for this 
type of experiment. 
The need for an improved fuel element modelling had always been pointed out by the 
Whole Core Accident Code expert group (WAC) of the Fast Reactor Safety Working 
Group (SWG). The WAC Group also recommended to develop further one of the exist­
ing fuel behaviour codes with the final goal of inserting it into the European Accident 
Code under development at Ispra. 
First contacts between Member States have shown that the experiments foreseen in 
HFR are of general interest. The discussions of the Fast Reactor Safety Working Group 
led to the same conclusions. 
Post failure phenomena 
Two types of in-pile experiments are foreseen referring respectively to a loss of heat sink 
at decay heat power level and a local subassembly accident. The main scope of these 
19.9.83 Official Journal of the European Communities No C 250/47 
experiments is the investigation of material relocation phenomena and of the in situ 
debris cooling potential. 
(a) Loss of heat sink 
Recent risk studies have shown that the loss of heat sink after reactor shut down 
cannot be ruled out completely. Situations can be imagined, e.g. the safe shut down 
earthquake for which the loss of heat sink has a probability in the order of 10-Va. 
An experiment is proposed to investigate the core behaviour under these circum­
stances, especially with regard to material relocation and fuel compaction. 
(b) Local subassembly accident 
During a local subassembly accident material motion and interaction determine the 
potential for propagation to adjacent subassemblies. Propagation may be caused by 
mechanical loading of the hexcan as a consequence of a pressure build up from a 
molten fuel/coolant interaction or by thermal attack. Thermal attack depends on a 
number of phenomena such as insulation of the hexcan wall by solidified fuel and 
the cooling capabilities of the damaged fuel bundle. 
Scope of the experiments proposed is the investigation of fuel element melt down, 
blockage formation and molten pool behaviour. 
Activities 
2.5.1. Transient fuel behaviour (point 1 .B.2.3) 
During a period of two years it is foreseen: (1) to define a detailed test programme to be 
performed in HFR which corresponds best to the needs of Member States performing a 
fast reactor development programme and (2) to perform a first series of experiments. 
The results from the programme preparation phase as well as the first series of experi­
mental results should provide the basis for a decision on the excecution of a more 
extensive irradiation programme which could be executed during the following years. 
2.5.2. Post failure phenomena (point 1 .B.2.4) 
(a) Loss of heat sink 
It is foreseen to perform a loss of heat sink accident simulation in a 37 pin cluster 
starting from a coolant flow reduction and ending by a slow fuel element melt 
down. Clad relocation and a fuel compaction are of main interest during the melt 
down phase. 
Preliminary studies have shown that the BR 2 reactor at the CEN Mol is well 
suited for such experiments. Local blockage experiments have already been per­
formed successfully in the same reactor (Mol 7C experiments). 
(b) Local subassembly accident 
Two series of experiments are envisaged, one in the BR 2, the other in the Scarabee 
reactor. 
In the experiments foreseen in the BR 2 reactor the whole sequence of a local acci­
dent will be investigated starting from a local blockage formation, up to the melt 
down of a subassembly, the molten pool formation and the possible thermal propa-
No C 250/48 Official Journal of the European Communities 19. 9. 83 
gation to adjacent subassemblies. The simulation of the whole accident sequence 
should, besides phenomenological data provide information on the evolution of 
the accident which is important for the assessment of the detection capability of a 
local fault. Experiments with a single fuel pin bundle and with two sections of two 
bundles are foreseen. 
Similar experiments are proposed in the frame of the Scarabee programme which is 
restricted to the study of molten pool behaviour and its cooling potential. The 
advantage of such a single phenomena experiment is that specific parameters can 
be varied more easily (e.g. U02, stainless steel, fission product rations) and the 
instrumentation can be more sophisticated. Consequently both experiments (BR 2 
and Scarabee) are complementary. 
Taking into account that: 
— the full execution of the proposed experiments would require a rather important 
amount of financial resources 
— technical feasibility studies are not completed for all experiments (BR 2, Scarabee 
partially) 
a two years preparatory phase is proposed during which in collaboration with national 
experts the details of the experiments will be further discussed. Where necessary addi­
tional feasibility studies will be carried out. 
It is planned to perform during this period two experiments in the Scarabee reactor and 
a first series of experiments in the HFR reactor. 
A decision on another experiment in Scarabee as well as on a continuation of the HFR 
and the BR 2 programme will be taken on the basis of further preparatory work by the 
end of 1985. 
Community contribution: 
4 800 000 ECU. 
2.6. FISSION PRODUCT TRANSPORT (Reference: point l.B.2.6. of the RAP) 
Objectives 
This programme unit addresses the description of radioactive material redistribution fol­
lowing severe accidents, the evaluation of containment, and the generation of the source 
terms required by codes used to calculate any ensuing offsite damage. Even if contain­
ment is rarely challenged realistic source terms are needed for risk analysis. Realistic 
internal distributions are needed for the design of containment and post accident moni­
toring systems. 
Realism implies avoiding excessive conservatism and taking into account inherent con­
sequence mitigating phenomena such as the reductions in chemical potential asso­
ciated with compound formation, adsorption and solution, and the loss of mobility 
associated with condensation and the agglomeration of aerosols. It also implies having 
available from other programme units or elsewhere realistic temperatures and descrip­
tions of how and when overloaded fission product barriers fail. On the other hand the 
desire for a more mechanistic approach to accident consequence analysis should not be 
confused with a need for high precision. A major challenge to containment must remain 
a rare event. 
The Safety Working Group and the Containment Working Group have examined mat­
ters related to fission product transport and containment since 1978, and commissioned 
studies to further define their work programme. 
19.9.83 Official Journal of the European Communities No C 250/49 
It is as a result of these studies that we emphase the value of even small delays between 
containment challenge and any failure. A pilot study of the importance of chemical 
interactions between fission products and other materials present in LMFBR is 
expected to further guide activity selection. 
In order to build most effectively on the technological base developed during the study 
of thermal reactors. The LMFBR programme necessarily emphasises activities that arise 
from the presence of sodium as a coolant, and sodium oxide as a carrier of fission prod­
ucts in the secondary containment. The value of fundamental thermo-chemical data, the 
application of phase composition analysis and the study of the more fundamental 
aspects of aerosol physics is common. 
This programme is designated to lead to an accepted code (or codes) for the description 
of in plan radioactivity distribution following a severe accident, a recommendable set of 
input data, and source terms for off-site accident consequence analysis. 
Activities 
— Definition of a unified set of thermodynamic data for the materials present follow­
ing melting of a LMFBR core during a HCDA for use in calculating fission prod­
uct release rates and the analysis of melt-containment compatibility problems. 
This activity will comprise theoretical analysis, experimental verification of phase 
diagrams and experimental verification of calculated vaporization rates for radiol-
ogically important species containing, for example, plutonium, higher actinides, 
iodine, caesium, ruthenium. 
Description of reactions between core melt and concrete will in addition require 
reaction rates. 
The experimental work performed to obtain equilibrium data will yield some 
kinetic data for well defined materials. Nevertheless provision has been made for 
the evaluation of existing data on uranium concrete reactions, the performance of 
additional experiments with urania-plutonia mixtures arid, to the extent deemed 
necessary, more complex mixtures. 
— Radioactivity distribution in the primary circuit of a LMFBR during and following 
a severe core accident; 
This activity covers the partition of activity between fuel, molten material (includ­
ing sodium) and gaseous phases such as bubbles or the cover gas. 
The desired output is a model describing the rate at which radioactivity leaves the 
primary circuit through any leaks that might develop, e.g. in the thick, rather com­
plex roof structure. Because certain leak paths may be long, account may have to 
be taken of repartition of radioactive species within the leak 
Existing models will be further developed as more experimental data becomes 
available. The aerosol codes to be validated (see the following paragraph) will need 
to be adapted to describe systems with moving boundaries in bubbles. Experi­
mental work on the transfer of radioactive species from hot bubbles to sodium will 
be undertaken, following review to ensure coordination with existing activities in 
Member States. 
— Radioactivity distribution in secondary containment. 
Effort will be focussed upon aerosols, because they are the dominant carriers of 
radioactivity, and the estimation of temperature and pressure changes during acci­
dents. 
No C 250/50 Official Journal of the European Communities 19. 9. 83 
The more precise estimation of temperature and pressure changes requires further 
modelling of sodium fires (covered in the programme unit 'Safety aspects of 
Sodium Technology') and the subsequent transfer of heat to available heat sinks. 
Experimental verification of sodium fire kinetics codes is foreseen in the above 
mentioned programme unit. Some of these experiments will provide data for verifi­
cation of aerosol codes already under development by Member States. 
The future development of aerosol codes is expected to include: 
— modules describing very dense aerosols; 
— attenuation in leakage paths; 
— re-examination of the fundamental equations used to describe agglomeration, 
gravitational settling and particle shape; 
— further experimental verification that changes in the composition of an aerosol 
following changes in the composition of the source are correctly modelled may 
be required when there has been an opportunity to review recent work. 
Aerosol codes for LMFBRS have many features in common with those used 
for LWR accident analysis. 
— LMFBR containment evaluation. 
This activity combines the results of the above activities with descriptions of con­
tainment structures, closures, materials and air-cleaning systems to provide the 
source terms needed for off-site damage assessment. 
The scenarios to be considered will include failure of containment systems, and 
structures when appropriate. 
Community contribution: 
2 900 000 ECU. 
2.7. MOLTEN MATERIAL MOTION AND INTERACTIONS (Reference point l.B.2.4. 
of the RAP) 
Objectives 
For LMFBR accidents leading to severe fuel damage the motion of molten material and 
its interactions with the coolant and structures have a strong influence on the sub­
sequent accident evolution. If for example sufficient quantities of molten fuel are 
expelled from the core region, during a severe accident, this may bring the reactor power 
to zero before larger parts of the core get involved. On the other hand if fissile material 
is driven into core regions having a higher reactivity value, the reactor power may 
increase and reach levels that lead to a whole core involvement and possibly to its 
desintegration. It should be noted that these scenarios are based on the assumption that 
the shut down systems are not intervening, therefore they have a very low probability. 
It follows that during all phases of an accident involving severe fuel damage, material 
motion must be taken into account. Considering the great number of rather complex 
phenomena coming into play a rather sophisticated mathematical modelling is required. 
19. 9. 83 Official Journal of the European Communities No C 250/51 
Amongst the phenomena to be considered energy, mass and momentum exchange 
between various components (steel, fuel, coolant, fission products) and phases (solid, 
liquid, gaseous), are of paramount importance. They influence for instance the transfor­
mation of thermal into mechanical energy which in its turn determines the loading and 
possible failure of core and containment structures. The attack of core structures by 
molten material is another cause for structure failure. 
During the early accident phases phenomena like freezing of molten material are impor­
tant as due to blockage formation fuel ejection from the core into the plenum may be 
stopped or reduced by the formation of blockages. 
Although a great effort in the Member States as well as in the US is devoted to the 
modelling of molten material motion and interactions, the results obtained so far are 
still not fully satisfactory from the point of view of a realistic description. Generally 
conservative assumptions are made which lead to very high mechanical energy releases 
during loss of flow and transient overpower type accidents. These very high energy 
releases are now considered as physically impossible. 
A more realistic description of these phenomena which is the main scope of the research 
proposed should provide the arguments for the elimination of an excessive conser­
vatism. 
Besides a direct contribution to research in this area, a Community action should also 
contribute to streamlining the on-going effort and stimulating the exchange of informa­
tion amongst all countries performing work in this field. 
Current work at the JRC and the activities foreseen in the next pluriannuel programme 
on EAC development, SIMMER code improvement and validation, FCI, plugging and 
freezing phenomena, post-accident heat removal, provide a valuable basis for a more 
extensive Community action. 
The expert groups WAC (Whole Core Accident Code) and CONT (Containment Load­
ing and Response) of the Fast Reactor Safety Working Group (SWG) both followed 
very closely current programmes in the field described and made proposals for future 
activities. 
Activities 
•— Definition of a unified set of thermophysical and thermomechanical data for 
materials relevant to severe accident analysis 
— Code development and validation for multiphase, multicomponent phenomena 
treatment 
— Experimental investigation of specific molten material problems, e.g. ejection from 
the core, interaction with structures. 
Community contribution 
6 000 000 ECU.