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THE ROLE OF NUCLEAR REACTOR CONTAINMENT IN SEVERE ACCIDENTS
Report by an NEA Group of Experts April 1989
NUCLEAR ENERGY AGENCY ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT Pursuant to article 1 of the Convention signed in Paris on 14th December 1960, and which cam: into force on 30th September 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed: - to achieve the highest sustainable economic growth and employment and a rising standard of living in Member countries, while maintaining financial stability, and thus to contribute to the development of the world economy; - to contribute to sound economic expansion in Member as well as non-member countries in the process of economic development; and - to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance with international obligations. The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, the Federal Republic of Germany, Greece, Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The following countries became Members subsequently through accession at the dates indicated hereafter: Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971) and New Zealand (29th May 1973).
The Socialist Federal Republic of Yugoslavia takes part in some of the work of the OECD (agreement of 28th October 1961).
The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC European Nuclear Energy Agency. It received its present designation on 20th April 1972, when Japan became its first non-European full Member. NEA membership today consists of all European Member countries of OECD as well as Australia, Canada, Japan and the United States. The commission of the European Communities takes part in the work of the Agency. The primary objective of NEA is to promote co-operation among the governments of its participating countries in furthering the development of nuclear power as a safe, environmen tally acceptable and economic energy source. This is achieved by: - encouraging harmonisation of national regulatory policies and practices, with particular reference to the safety of nuclear installations, protection of man against ionising radiation and preservation of the environment, radioactive waste management, and nuclear third party liability and insurance; - assessing the contribution of nuclear power to the overall energy supply by keeping under review the technical and economic aspects of nuclear power growth and forecasting demand and supply for the different phases of the nuclear fuel cycle; - developing exchanges of scientific and technical information particularly through participation in common services; - setting up international research and development programmes and joint under takings. In these and related tasks. NEA works in close collaboration with the International Atomic Energy Agency in Vienna, with which it has concluded a Co-operation Agreement, as well as with other international organisations in the nuclear field.
Public en frar.Yjis sous Ic titrc :
I.E ROLE DU CONFINEMENT DES REACTEURS NUCLEA1RES EN CAS D'ACCIDENT SEVERE
©OECD, 1989 Application for permission to reproduce or translate all or part of this publication should be made to: Head of Publications Service, OECD 2, rue Andre-Pascal, 75775 PARIS CEDEX 16, France. FOREWORD
In November 1986. the Senior Group of Experts on Severe Accidents of the NEA's Committee on the Safety of Nuclear Installations (CSNI) was invited to examine the role of containment in severe accidents and to report to the Committee on the outcome of its discussions.
The Senior Group's discussions have strengthened and supplemented with more detail the conclusions reached in its previous report, "Severe Accidents in Nuclear Power Plants" (published in May 1986). It is now widely recognised that containments should play a major role in the management of the severe accidents. No fundamental shortcomings calling for radical change have been identified in existing designs.
Current relevant R&D activities are reasonably extensive and show an awareness of major issues; the Senior Group wishes to emphasize that it has not identified major shortfalls in existing research. Nevertheless it regards continuing research by OECD Member countries as vital, not least because of the need to base decisions on a realistic approach rather than on limiting (so-called "conservative") cases which might lead to inappropriate procedures.
The experts who prepared the report are listed at the end of this volume.
3 Table of Contents
EXBCUTIVE SUMMARY 5
PART I - CURRENT DESIGN STATUS 8
Containment Structural Capabilities 8
Containment Loading in Severe Accidents 10
Direct Containment Heating (DCH) 12
Hydrogen Combustion 13
Molten Core-Concrete Interactions and Debris Bed
Coolability 14
PART II - ACCIDENT MANAGEMENT POSSIBILITIES 17
PART III - GENERAL GOALS FOR ACCIDENT MANAGEMENT 19
Reducing the Potential for Pre-Existing Openings 20
Reducing the Potential for Early Failure 21
Reducing the Potential for Late Failure 21
Filtered Containment Venting Systems 22
Long Term Control 23
Conclusions 24 LIST OF AUTHORS 27
4 Executive Summary
1. One of the principal conclusions in the report of the Senior Group of Experts on Severe Accidents (published May 1986) was that containments* in general have great potential to mitigate the consequences of a severe acci dent. They are not designed specifically for that purpose, but in practice the specifications of the basis to which they are designed and constructed enable them to handle events far beyond design. The TMI-2 accident exempli fied how even an accident severe enough virtually to destroy the reactor core resulted in only negligible radiological consequences to the external envi ronment because of the degree of control made possible by the containment structures.
2. However, the Senior Group pointed out that realising the full potential value of the containment to the control and management of a severe accident would require a concerted effort involving diverse experts normally accustomed to operating independently of each other. At its plenary meeting in 1985 CSNI appointed a Task Force on Containment Performance to review the possibilities and needs, resulting in a detailed and wide ranging technical survey.
3. Other activities inspired or encouraged by CSNI have produced relevant material, notably Specialists' Meetings such as that on Core Debris - Concrete Interactions held in Palo Alto in September 1986. There has also been a short debate centred on the USNRC's "Reactor Risk Reference Document" (NUREG 1150). CSNI in their plenary meeting in November 1986 invited the Senior Group of Experts on Severe Accidents to address the question of the role of the containment.
4. The Senior Group has reviewed developments in research and operating practice which have occurred in the three years or so which have elapsed since it was formulating its Severe Accident Report. Although nothing transpired from this review suggesting that the views expressed earlier needed radical modification, the question of containment has been examined in the somewhat greater depth and detail made possible by extensive use of such material as that produced by the Task Force.
* In its 1986 Report the Senior Group defined "containment" as a structural envelope which completely surrounds the reactor system and is designed to hold the releases from design basis accidents with little or no release to the environment. The term is used in its broader sense to include associated leakage paths and buildings which contain the releases of severe accidents.
5 5. To provide an adequate background for its conclusions the present report begins with a summary of the general viev on containment and the more important phenomena relevant to it in severe accidents. The resultant loads, mechanical and thermal, are then considered together with the capability of the containment to vithstand them and the modes in which various failures might lead to loss of containment integrity. Then possibilities are reviewed in the light of currently developing ideas, of developing the potential of the containment for minimising the effects of severe accidents. The Group con fined its discussion to existing plants, excluding future designs.
6. Generally, a LWR containment is designed on the basis of conservative assumptions to maintain a high standard of integrity (leak tightness) with a margin beyond accidents (loss-of-coolant accidents) which, although unlikely, must be included in the design provision and which may entail the release of significant amounts of radioactivity from the primary circuit. This function of preventing release of radioactivity to the public has to be performed de spite the pressures and temperatures associated with the release of primary fluid. There are various ways to achieve that general objective. Some de signs use large free containment volume whereas others employ quenching devices which reduce the free volume and structural strength required. No universal prescription is possible in view of the range of variants. Never theless broad conclusions can be drawn for existing reactors, stressing that they are to be interpreted in the light of the overall detailed concept of any individual design, taking account of ancillary equipment provided, such as coolers and quenching devices in addition to the size and strength of the main containment structure.
7. Beyond the purpose which largely determines the containment specifica tion there are other duties often expected of the containmpent. These include a capability to withstand external loads, such as storms, earthquakes and mis siles. Whilst such considerations play a part in determining to some extent the detailed strength of the containment they are really part of the design basis and as such are outside the scope of the present report.
8. In addition, many safety specialists have stressed a vaguely-specified but significant benefit conferred by containment in providing an extra measure of protection which can compensate for any inadvertent omissions from the list of initiating events addressed in the design safety case. The role of contain ment in severe accidents can be seen as an extension of this idea of "defence in depth".
9. When considering the containment as part of the procedure for managing a severe accident its primary function of confining radioactivity released from the primary circuit must always be given priority and its design, con struction and operating rules must reflect that. However that need not pre clude the possibility, in extreme circumstances, of extending the range of available options, giving the operators a greater degree of flexibility in managing the accident. In that context the possibilities proliferate so wide ly as a postulated severe accident progresses that precise prediction of the course of events, and of the probability of successive discrete plant states is not possible. Therefore actions should be based on "best estimates" of the current state of the plant and of the most likely next stage of development of the accident. Again, because of the high quality of the designed safety pro visions a severe accident at such an advanced stage is a very improvable event.
6 As a consequence many hold the view that a rigorous cost-benefit analysis is unlikely to be effective in considering what additional measures (such as filtered venting) should be adopted. Clearly cost considerations play a part but so far those countries vhich have decided to adopt such measures consider that they do not incur a major addition to the overall cost of the plant.
10. The Senior Group concur vith the view expressed in the report 75-INSAG-3 (Basic Safety Principles for Nuclear Power Plants, IAEA, Vienna 1988) that severe accident management and mitigation measures have the potential to re duce by a factor of at least ten the probability of large off-site releases requiring short term off-site response. The remainder of this report examines in more detail hov, and to what extent, such reductions can be achieved.
11. The existing capabilities of current containment designs to withstand various loadings are reviewed and compared with conditions which may arise in severe accident conditions. Next some possibilities for enhancing the ability of the plant to reduce by accident management the risk of significant off-site consequences are examined. Finally, current R&D topics of particular signifi cance are identified and attention is drawn to the directions in which future endaavours may best serve the effective management of containment capabilities in severe accidents.
7 Part I
CURRENT DESIGN STATUS
CONTAINMENT STRUCTURAL CAPABILITIES
12. Containment design criteria are based on a set of deterministically derived challenges. Those for pressure and temperature challenges are usually based on the design basis loss-of-coolant accident; criteria based on exter nal events such as earthquakes, floods, and tornadoes are also considered. The margins of safety provided by such practices have been the subject of consid erable research and evaluation, and these studies have shown the ability of modern containment systems to survive pressure challenges veil beyond design levels. Because of these margins, the various containment types presently utilised in the OECD Member countries have the capability to withstand, to varying degrees, many of the challenges presented by severe accidents even though not designed primarily for that purpose. For each type of containment, however, there remain failure mechanisms which could lead to either early or late containment failure, depending on both the accident scenarios involved and the containment types.
13. The ways in which failure of Reactor Containment Buildings (RCBs) might jeopardise their task of preventing release of fission products to the environ ment are:
1. Leakage at penetrations such as personnel and equipment entries, and cable and other service entries.
2. By-pass through inadvertent openings.
3. Leakage caused by internal overpressure exceeding the strength of the RCB.
4. Melt-through of the RCB base by molten reactor core material.
5. Leakage caused by structural failure due to extreme external events such as earthquakes, aircraft crashes or plant-generated missiles.
14. In most cases the standard design approach has been dominated by the need to resist loads generated by an accident involving loss of primary cir cuit integrity. The application of probabilistic risk analysis has focussed attention on the capacity to withstand overpressurisation. Analyses of con tainment performance for PVRs under conditions beyond the design basis have demonstrated significant overpressure potential in many existing designs.
8 15. None of the design basis accident scenarios involve very rapid appli cation of load to the RCB structure. The conditions applying are therefore quasi-static. Dynamic loadings, including fluid jet impingement, and direct heating or rapid deflagration of hydrogen pockets following severe degraded core accidents, may represent a significant threat to a RCB.
16. Experimental testing using scale modelling techniques has be«~n used ex tensively and successfully to predict the performance of prototype structures over a wide range of engineering disciplines. The use of such techniques to predict reactor containment building behaviour under normal and accident load ing conditions is, in principle, the most reliable technique available. The chief disadvantage of the technique is the very practical one of adequate re presentation of the structural components if very small scale models are employed. With all-steel structures these problems are ameliorated with the least difficulty, but with concrete or composite structures there are practi cal limits to the thickness of concrete and the placement of reinforcing bars and prestressing tendons.
17. Currently a range of methods are available, from the early force bal ance equilibrium calculations to the more intricate finite element analysis made possible by the latest computer developments. Each has its advantages and limitations but present indications, in the light of recent experimental evidence, are that whilst their combined use can give a good indication of the behaviour of composite steel and concrete structures the analysis must be complemented by the results of experimental investigation.
18. The failure mode of a RCB is an important factor in assessing risk. The behaviour of the structure between design pressure and failure pressure is difficult to predict, although the prediction of the pressure at which failure occurs can be made with reasonable accuracy. The behaviour of the RCB close to the failure pressure, and the mode of failure, are all-important since they determine whether the structure fails suddenly and explosively or gradually with a progressively increasing leak and the maintenance of global structural integrity. Containment design has a major influence on failure modes, and there is some evidence that failure mode can be estimated for particular types. The strains induced in steel liners may depend very strongly on their struc tural interactions with the concrete, and in particular on the effects of liner-to-concrete anchors, pene .rations, and discontinuities in geometry. Further work is certainly necessary before accurate predictions of failure mode can be made for this mos'. common type of RCB.
19. The effects of external blast waves, with accident overpressures up to about 0.5 MPa, on concrete structures can be very adequately predicted using well validated finite element methods. However present calculation techniques cannot reliably predict the effect of a high intensity blast wave of a type which might be produced by a large hydrogen explosion.
20. In general, reactor licencing requirements in OECD Member countries impose the need for the reactor containment building to be able to survive a simultaneous LOCA and a 'safe shutdown earthquake" (SSE). Satisfying this requirement normally results in the provision of additional reinforcement in specific areas such as at the wall-basemat junction and in the basemat itself. This additional reinforcement provides some additional strength against the overpressure condition.
9 21. Assessment methods for concrete structures are based on finite element analysis of the reactor containment building when a specified excitation, typical of earthquakes experienced in seismically active regions of the world, and with stipulated peak accelerations dependent on the perceived typical earthquake experienced in the region on which the RCB is to be constructed, is applied to the structure. The degree of validation for the finite element codes used for such calculations is somewhat less than ful1y adequate, but the margins to failure are generally large.
22. Analysis of the behaviour of steel containment structures can also be achieved using finite element methods, and with confidence in the assessment of the structure's capability.
23. In terms of structural damage, crashing aircraft represent the major design basis hazard but impact probabilities are very site specific and are therefore not generally applicable.
24. The reactor plant itself is capable of generating energetic missiles from failures in rotating machinery, pressure circuit bursts and pipe whip events. It is also possible in some power plant configurations for machinery failures adjacent to the reactor containment building to produce hazardous missiles.
25. The ability of a concrete structure to resist perforation by an im pacting missile can be confidently predicted using methods well validated by experiments. The missile energy remaining after perforation is also calcula ble so that its influence on sensitive items is therefore assessable. Sub- perforation damage states are less well predictable in extent and detail, but these present much lower accident hazards. Secondary missiles formed as a consequence of scabbing are generally of low energy. Where equipment is sensitive to such low energy missiles, all presently available experimental evidence indicates that a relatively thin steel plate (about 3 mm) anchored to the distal face of a reinforced concrete wall will retain scabbed material completely and will thus eliminate secondary missiles. (Injection and subse quent conflagration of aircraft fuel is the only significant risk from sub- perforation damage.)
26. RCB behaviour due to dynamic forces resulting from jets issuing from breaks in steam lines is difficult to predict with any certainty. Imposed load diminishes rapidly due to divergence of the jet. The effects of local heating of a RCB structure due to high temperature jet impingement are also difficult to assess. Softening of steel structures will occur but to a rela tively small extent, and in the limiting case reinforcing bars and prestress- ing tendons will also be softened, but, in general, temperatures likely to be reached are relatively low and the consequences relatively unimportant for a containment building.
CONTAINMENT LOADING IN SEVERE ACCIDENTS
27. The key phenomena and processes important to the evolution of a severe accident and which can have an important effect on the containment behaviour are, in general, generic in nature with some variations depending on specific
10 reactor types (eg. PVRs, BVRs), other unique design features and also contain ment design configuration. They determine the containment loading mechanisms and associated vulnerabilities and, by their nature, suggest accident manage ment strategies for either eliminating or delaying failure; for example, proce dures or procedures coupled with hardware additions such as filtered-vented systems, at least for sequences involving slow overpressurisation of the con tainment. Even though the primary challenge to containment arises as a conse quence of primary system failure and the thermal-hydraulic-material interac tions following release of the molten reactor core debris into the containment space, certain containments can also be challenged as a consequence of the interactions taking place in-vessel before complete primary system failure; for example by the generation and uncontrolled conflagration of large amounts of hydrogen. It is for this reason that actions may be taken to mitigate specific challenges, such as inerting BVR containment atmospheres and install ing systems for the controlled burning of hydrogen.
28. The understanding of those phenomena and their range of behaviour is important to taking advantage of existing plant capabilities and exploring additional accident management strategies for the different reactor/containment types.
29. The resulting severe accident phenomena are capable of generating a much higher level of loading than that envisaged in the design basis and have their origin in a variety of initiating events and sequences generally in volving multiphase flow and heat transfer processes. The phenomena and the specific mechanisms involved in these processes have been under intensive study worldwide, especially following the TMI accident. They may be summa rised as follows in terms of nomenclature employed in WASH 1400 for PVRs:
Alpha: Steam explosion or large scale in-vessel molten reactor core-water energetic interaction with potential to cause early containment failure via energetic missile penetration of the containment.
Gamma: Hydrogen deflagration/detonation. This could lead to con tainment failure in cases of detonation and also in certain limit conditions in cases of deflagration; it is more or less important according to the type of reactor and contain ment design.
Early Rapid containment overpressurisation early in the accident. delta:
Delayed Usually gradual overpressurisation at a late stage of the delta: accident.
Epsilon: Basemat melt-through via its decomposition from contact with high temperature core debris. It is also recognised that earlier deterioration may occur by direct attack by molten debris on steel components of the boundary, depend ing on the particular design and sequence.
Beta: Failure to isolate the containment.
V: Containment bypass (or interfacing system LOCA).
11 In addition to the failure modes listed in WASH 1400 leakages induced as a result of high pressures and temperatures are to be considered.
There are indications of a growing consensus that nowadays the alpha, gamma and early delta failure modes may in principle be considered of low condition al probability, especially where countermeasures are applied. For example, direct heating may possibly be avoided if there is an option for depressuris- ing the primary system prior to its failure. Detonation may be avoided on some existing reactors by using preventive measures such as igniters with a reasonably wide scope of application. The delayed delta mode is seen in some countries as susceptible to control by filtered venting, possibly in associa tion with sprays fed from back-up water. The epsilon mode is considered relatively likely in the case of dry contact between core melt and concrete. A dry basemat is assessed to have been penetrated in several day? or even weeks; the release of fission products to the outside atmosphere is howeve.- relatively low as compared with other release paths. Although large scale tests have not been performed it is possible that deep water pools of suffi cient cross section area would cool down injected debris or melt. The effectiveness of limited amounts of water to cool the core melt has not yet been studied in much detail. Although the beta mode is generally viewed as significant, much progress has been made by paying particular attention to the possibility of small pre-existing leaks and to the confinement function that may be provided by adjoining buildings to which the radioactivity leaks. Similar considerations apply to the V mode which, although still regarded as significant, is now considered somewhat less so than formerly. The probabili ty of failure of steam generator tubes is considered as significant either as a cause or as a consequence of core damage; appropriate management of the accident may avoid or mitigate its consequences.
30. The phenomena associated with molten core-concrete interactions, debris bed coolability - including steam spikes, and hydrogen deflagration/detonation have been selected for more detailed treatment below. These phenomena have been selected because of their significance, given the occurrence of a severe accident. Also "practical" steps can be visualised, wherein the existing con tainments, coupled with simple hardware/procedural additions, can be utilised to reduce the contribution of those phenomena to the likelihood of significant release, possibly eliminating it altogether.
31. Other sequences are also considered for which the conditional probabili ty of occurrence may be higher. However, if they do lead to containment fail ure it would be after a long delay giving more scope for accident management to bring about a significant mitigation of consequences.
DIRECT CONTAINMENT HEATING (DCH)
32. This process is postulated to arise from an accident sequence which de velops while the primary circuit is at high pressure. Failure of the primary circuit integrity would in such circumstances be most likely to occur when the core was extensively molten. If the breach in the primary circuit occurred at a part of the boundary retainiiv molten debris the result would be a massive expulsion of very hot liquid material driven by the primary circuit pressure. The consequent dispersal of finely divided debris may lead to rapid heating of
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MOLTEN CORB-CONCRETE INTERACTIONS AND DEBRIS BED COOLABILITT
36. If hot core debris penetrates through the bottom of the reactor pressure vessel, it will come to rest on the concrete floor of the containment in most designs.
37. The principal ways in which containment integrity may be lost due to the core debris-concrete interaction are:
- Structural failure due to overpressurisation or thermal stresses.
- Leakage at seals or penetrations, induced by overpressurisation or high temperature.
- Penetration through the concrete basemat into the underlying soil or rock.
38. In a core-concrete interaction heat is transferred primarily to con crete which decomposes, absorbing the latent heats of dissociation and of melting while giving off gases and slag. The gases are mainly water vapour and carbon dioxide. If the molten core debris contains un-oxidised zirconium, iron, nickel or other metals, as it may, these will be oxidised by the water vapour and carbon dioxide with the evolution of hydrogen and, under some con ditions, carbon monoxide and methane in smaller quantities. Heat will also be produced. The addition of non-condensible gases to the containment atmosphere will raise its pressure. Calculations assuming specified geometry, layer depth, decay heat and melt composition indicate that the rise in pressure, superimposed on the pressure rises due to other aspects of the hypothetical accident, may be sufficient to cause structural failure of the containment shell. Decay heat, possibly enhanced by oxidation reaction heat, may be suf ficient to weaken structural portions: of the containment shell. Heat radiated from core debris may be sufficiently intense to cause weakening of concrete pedestals or load-supporting vails. The interaction could result in penetra tion of the debris through the concrete itself.
39. A particular problem affecting ceitain reactors is risk of contact betveen the containment contaminated sump water containing practically all the dissolved volatile fission products and the very hot core debris. Energetic contact reactions might result in release into the containment of non- negligible quantities of fission products, in a situation where pressure in the containment may already be very high with respect to the ultimate mechani cal strength defined for the structures concerned. However this may be only an intermediate stage towards a final stable configuration in which all debris is covered by water.
H 40. The composition of the concrete is a significant factor in determining the quantity of gas evolution, the resulting overpressurisation and the rate of penetration by the debris. The overpressurisation effects are most notice able in relatively small containments, as are the thermal effects of the core debris/concrete interaction.
41. The most significant aspect of the core debris/concrete interaction issue is its relationship to the source term. The issues of containment structural damage due to overpressurisation or overheating, and of basemat penetration by the debris can be addressed by applying existing analyses and engineering judgement.
42. Some of the codes used by researchers in this area appear to under- predict the axial erosion of concrete when compared with experimental data from the BETA facility, at least at early times when, moreover, the predic tions of the codes vary considerably. However, at later times the total volume of concrete ablated and hence gas evolution as predicted by the various codes are generally similar. This supports the conclusion that at later times core-concrete interactions are controlled by a simple steady state balance of internal heat generation and the available heat loss mechanisms from the core debris.
43. Arguments can be adduced that for certain designs in which there is room for debris to spread out and to be submerged in water the debris will be cooled so that gas evolution is effectively limited. The efficacy of attempts to control core-concrete interactions by applying water seems to depend sig nificantly on the amount of water available and on the nature of solid crust formation at interfaces between water and melt. Some additional work, perhaps including experiments, is needed to assess the adequacy of that strategy in preventing concrete basemat penetration. It must also be borne in mind that if the application of water is effective in cooling the debris the heat there by conveyed by steam has to be rejected by the systems for condensing steam in the containment or other means. If the debris is not effectively cooled the permanent gases generated would cause overpressurisation but that could be countered by venting the containment.
44. In conditions where it has not proved possible to re-establish an efficient heat sink inside the containment so as to remove residual heat, the effective mode of containment failure will depend on the relative importance of the physical conditions in the particular circumstances obtaining. For instance, in the absence of water, basemat penetration would appear to be very probable whilst the internal pressure buildup, mainly due to non-condensible gases alone, will be slow. On the other hand, the injection of water to cool the core debris may prevent or delay full penetration of the basemat. The residual heat would however form large quantities of steam which would have the drawback of considerably accelerating the pressure buildup in the contain ment depending on the quantity of core debris falling on the basemat.
45. In summary, accident management strategies can be devised to deal with the potential consequences of core-concrete interactions. They may arise unless attempts to utilise water for cooling core debris are successful, particularly if the cavity configuration favour? core-concrete interaction. In that event containment failure from overpressurisation could be prevented by venting strategies. The steam spike from the cooling process could be mitigated either by using natural processes or by introducing "condensation"
15 strategies possibly in association with the venting of the containment. Core- concrete interactions with or vithout the presence of water need continuing attention both analytically and experimentally, especially the "augmented" source term which could be generated from the variety of reactions with the metallic components of the molten debris. However, the Senior Group still consider that on balance the addition of water is to be strongly recommended.
16 Part II
ACCIDENT MANAGEMENT POSSIBILITIES
46. In the two previous sections the general capabilities of containments to withstand mechanical and thermal loads and the nature of containment load ings in severe accidents have been reviewed. Although containment capabili ties are determined primarily by the requirement to deal with design basis accidents it is clear that significantly greater loads could be accommodated. The purpose of accident management is to secure maximum benefit irom that mar gin of strength.
47. The release of the inventory of the primary coolant system results, in general, in an initial pressure increase followed by a decrease. There are various designs for achieving the aim of containing the release.
48. Large dry containments are in principle passive systems for a consider able time after the release. The subsequent pressure decrease is effected by spr-^s within the containment or by heat removal in other heat exchangers. Containments with a small volume need a prompt capability to condense the steam either in a water pool or in ice beds through which the steam/air mix ture has to be directed. In general containments do not need electrical power for some time; many designs can absorb the decay heat safely for hours or even days. However, if the capability to absorb decay heat is inadequate the pressure could soon exceed the design value.
49. For some designs water exists beneath the reactor pressure vessel and contact with the melt may result in a rapid steam production. If, however, the vater pool is large and the water subcooled, the melt might be cooled down completely without severe pressure production.
50. One of the main potential loads for the containment could result from a hydrogen burn, if oxygen is available. The resulting pressure depends on whether detonations or deflagrations may occur, leading to very high loading. In principle the risk could be countered by post-accident inerting of the containment atmosphere. However, very large quantities of inert gas (eg. at least 100 tonnes of nitrogen) would be needed, presenting significant problems of storage and injection. The disturbance to the pressure and temperature of the containment atmosphere is not easy to predict with confidence and could result in replacing steam with nitrogen,which is less effective an inhibitor than steam. In any event the hydrogen would remain and could present a hazard at a later stage. Thus the advantage secured by inerting is not clear for large dry containments.
51. Containment integrity may also be threatened by long term pressure and temperature build-up and by a possible melt/concrete interaction if adequate coolability of the melt cannot be achieved. These loads would threaten con tainment integrity within anything from less than a day to several days depend ing on the proportion of decay heat producing steam.
17 52. Finally it has to be borne in mind that even when the plant has sur vived successfully the dangers of inadvertent leakage, and early and late structural failure, there will be a long-term need to maintain control. In favourable cases restoration of normal pover supplies and services may enable this to be done relatively easily. Even then, however, management will be required to ensure that continuing attention is paid to anticipating in good time possible deterioration of equipment and structures. In less favourable circumstances there may be a need for urgent and more or less major provisions involving, for example, additional power supplies, services, equipment and additional structural work.
53. In seeking a management strategy the first task to be undertaken is the analysis of how the containment can cope with the loads from a severe accident. This task is plant specific, in general. In addition to realistic assumptions and calculations expert judgement is sometimes necessary to assess the capabil ities of the containment.
54. The containment function is jeopardised by pre-existing openings (beta- mode) in the containment boundary or by containment bypass. Some progress has been made during recent years in avoiding pre-existing openings.
55. An early containment failure usually results in an immediate high re lease of radioactive fission products to the environment. Moreover an un controlled leakage cannot usually be closed by normal measures and thus allows a continuing release of fission products. Loads early in the severe accident sequence may arise in various ways:
- It is generally considered - although not completely established - that a steam explosion (alpha-mode) threatening containment integrity is unlikely.
- The loads from a hydrogen deflagration or detonation (gamma-mode) are very sequence- and plant-specific and no generic assessment is applicable to all cases.
- Loads from rapid steam production resulting from a core melt/water interaction are currently assessed as not to result in a threat to containment integrity.
Research during recent years has indicated that rapid heat addition to the containment atmosphere ("direct heating") could, for some designs, result in containment failure. This conclusion is sensitive to the assumptions and method of analysis employed and more research is under way to study this process.
56. The most likely routes to late containment failure are pressure and temperature build-up due to evaporation of water or the production of non- condensible gases from a basemat melt-through. However, the fission product release from a late failure is usually very much less than for early failure and more time is available for appropriate management actions.
18 Part III
GENERAL GOALS FOR ACCIDENT MANAGEMENT
57. The following chapter suggests appropriate goals and discusses the procedures to attain them. The design-basis accidents could be limited to a few sequences because they are of "umbrella"-type and include some margin. The procedures to cope with design-basis accidents are mostly event-oriented but am increasingly supported by symptom-oriented procedures. Because of the variety of possible severe accident sequences symptom-oriented procedures are appropriate for tham in conjunction with knowledge of the states of systems, components and structures. Furthermore they should be assessed on a realistic rather than a "pessimistic" basis. Specification of a goal for an acceptable reduction in the risk of fission product release is a matter for the competent authority in each member state. In dealing with high potential consequences correspondingly low probabilities are appropriate, but very low numerical values of probability do not always carry conviction because of the wide margins of uncertainty associated with them. Nevertheless the Senior Group are of the opinion:
i) Acceptable releases must be coupled at least in some way with a judgement of the probability of occurrence.
ii) There are reasonable and practicable ways to attenuate releases in severe accidents and they should be implemented.
In the following section the beneficial aspects of certain general plant design and operation features in reducing risk of fission product releases are considered.
58. Existing instrumentation is designed to identify the status of the plant during normal operation and design basis accidents. In the face of the specific loads and requirements imposed during severe accident sequences the existing instrumentation may not be adequate and may have to be improved and perhaps supplemented. The main requirements are as follows:
- The instrumentation should cover the foreseeable ranges of the variables to be measured.
- The instrumentation has to survive the loads and environment during severe accident sequences, eg. high pressures and temperatures, loads from a hydrogen burn and high radiation levels.
- The instrumentation should remain intact for long periods because repair or replacement might not be possible in high radiation levels.
19 - The instrumentation should cater for such major variables as pres- sure, temperatures and composition of the containment atmosphere as well as the state of valves and of injection systems. The knowledge of radiation levels at different locations would be of helo for some accident management measures.
In addition to appropriate instrumentation, computer aids to the operators should be developed and installed to assist liagnosis of the situation. Their main objectives should be:
- Description of the status of the plant;
- Providing a basis for the identification of the history of the event and for prognosis of its future development.
59. Appropriate management actions to mitigate the effects of pre-existing openings and early or late failure can be performed with existing equipment, possibly enhanced by additional measures. Because the possible measures are plant- and sequence-specific no generic recommendations can be given.
60. In order to identify and recommend appropriate accident management actions, performance goals for the containment need to be specified. The following are proposed:
1) Reduce the potential for pre-existing openings in the containment boundary and between wetwell and drywell for containments with pressure suppression systems.
2) Reduce the potential for an early containment failure.
3) Reduce the potential for late containment failure.
A) Develop the ability to foresee long term requirements and the flexibility to meet them.
REDUCING THE POTENTIAL FOR PRE-EXISTING OPENINGS
61. Existing containment leakage requirements are normally intended to limit public radiation doses to those prescribed by regulations and are based on specific plant and site characteristics. Accident analyses should demon strate that the allowable leakage rates will not result in exceeding the site boundary radiation dose criteria under postulated accident conditions. Peri odic integrated containment leakage tests and local leakage tests are reauired during the operational life of the plant in order to demonstrate continuing compliance with the allowable leakage rates. Operating experience on contain ment leakage testing indicates that leakages often exceed the technical speci fications. "As found" leak values are often greater than the "as left" values after the preceding test, and the "as found" values mainly depend on the time interval between two consecutive tests.
20 62. Generally, abnormal containment leakages are assumed to be caused by defects in the Containment Isolation System. In fact even if the Isolation System operates correctly, abnormal leak paths might exist because of imper fect tightness of valve or airlock seals. Defects of the Containment Isola tion System may be regarded as pre-existing openings in the containment struc ture at the time vhen a severe accident is postulated.
63. Efforts are currently under way to derive improved risk estimates. It is thought that the use in risk analysis of more dependable probabilistic data on pre-existing openings could provide support to specific actions for signifi cantly improving the containment performance during severe accidents.
64. Pre-existing openings should be easily detectable and studies should be undertaken on means to close the openings. The possibility of long term under pressure in the containment after closure should not be overlooked. Leakages between the wetwell and dryvell may arise from open doors, valves left or stuck open or leakages in the condensation pipes. Pre-existing openings can be detected by global low pressure containment leakage tests performed during reactor operation. Such tests are performed on a regular basis in Belgium and in France, and it has been shown that they are able to detect holes larger than 0.4 cm2.
REDUCING THE POTENTIAL FOR EARLY FAILURE
65. The causes for an early failure depend strongly on the design of the containment and the sequence. They include:
- Overpressure failure due to the production of non-condensibles for small containments, which could be avoided by a venting system.
- Heat-up of the atmosphere due to melt and possible related reactions (direct heating), which could possibly be avoided by decreasing pres sure in the coolant circuit before ejection or limiting the overpres sure generated by impeding communication between the pressure vessel cavity and the upper containment.
- Melt-through of steel walls by melt or debris, which could be pre vented by fitting additional protective shields and, for less violent expulsion, by cooling the debris.
- Temperature and pressure increase from a hydrogen burn or hydrogen detonation which could be countered by pre-inerting for small volume containments or by providing catalytic devices or igniters to re- combine hydrogen.
REDUCING THE POTENTIAL FOR LATE FAILURE
66. Two effective measures against an overpressure failure are the use of the existing containment spray system possibly augmented by independent sources
21 of water and power supplies fir its transport, and a venting system. Flooding of the cavity below the pressure vessel can also be an effective measure to condense steam. Basemat melt-through can be avoided if the melt can be cooled by water or special heat exchangers.
FILTERED CONTAINMENT VENTING SYSTEMS
67. Pressure relief by a predetermined leakage (vent-system) would be an appropriate mesure to avoid an overpressure failure of the containment, and then to provide the operator with the means of controlling the containment function, both above and below ground, until the end of the accident, whatever its severity, and to limit the level of releases into the environment. It has already been decided that this measure will be installed in PVRs and BURs in France, the Federal Republic of Germany, Sweden and Switzerland, and variants of it in some of the reactor types utilised in the United States.
68. The required venting capacity depends on the sequence, but it is clear that very sharp pressure rises, eg. from hydrogen burn, melt/water interaction or melt/atmosphere reaction, cannot be significantly reduced by a vent-system.
69. Nothing must detract from observance of the design rules for the whole installation, nor, in the event of implementation of these systems, cause worsening of the accident. The problems concerning safety aspects directly related to the use of the system must be clearly identified; each one has then to be solved according to the specificity of the system. The design experience acquired in the countries already engaged in this way, indicates so far that it is manageable.
70. Some specific problems:
- It is important that a vent-system should not create a weak point in the containment boundary; therefore it must be designed so that no failure should occur below the failure pressure of the contain ment. In addition, the valves should have the capability to close against the maximum pressure in the containment.
- Whether and when to open the vent-system depends very much on the sequence, vent capacity and water addition to the containment. In general, the vent-system should be operated above the design pres sure but well below the failure pressure. The available tine, the required reliability of the system and the access to the valves will finally decide whether the venting should be initiated automatically or manually.
- Air, possibly hydrogen and steam will be released through a vent- system. It is necessary to analyse what underpressure in the con tainment would result after the condensation of steam, which results from heat transfer to the environment or after the start of a con tainment spray system. Addition of water to the containment sump will decrease steam production and also compensate for released water and so extend the availability of heat removal capacity.
22 LONG TERN CONTROL
71. Control of the installation in the long term is defined as the quasi- steady state situation vhich should exist after the intermediate efforts, including possible use of any filtered-vent system, have gone to completion. It is envisaged that the time-span of long term control is from a fev days to a fev months. The features to be emphasised in long term control are:
a) Heat Rejection:
Systems to transport heat from the vessel (if intact) or containment to the heat sink (if vessel breach has occurred) must be reactivated. This may involve transport of highly contaminated vater outside of the reactor building to pumps and heat exchanger.
b) Containment Atmosphere Control:
The pre-accident containment atmosphere is primarily non-condensible gases. During and immediately after (especially if venting has occurred) there may be a replacement of non-condensible gases by vater vapour, in vhich case long term containment atmosphere conditions could result in condensation leading to an under-pressure. Care must be taken when restoring containment cooling, in connection with isolation of any vent system, to avoid the implosion mode of containment failure.
In addition, restoration of containment cooling could result in steam condensation so that hydrogen combustion becomes more of a problem, and should be taken into account.
c) Criticality:
In most cases it will be necessary to assure high concentrations (several thousand ppm boron) in circulating water in the RCB to assure subcriticality of the core, whether intact or disassembled.
d) Water Chemistry:
In the long term the pH of the RCB circulating water should be con trolled to the neutral or slightly basic point, to avoid re-evolution of iodine.
e) Maintenance and Equipment Survivability:
In the long term there will be, possibly, exposure to high radiation of active components involved in heat rejection, and passive components involved in measurements and observation. Consideration of maintenance of such components, or innovative use of alternative systems, such as robotics, should be part of the long term control element of accident management.
23 f) Post-Accident Monitoring:
There must be continuing provision to monitor dose levels, radiation releases and the inventory of activity in the various parts of the plant.
CONCLUSIONS
72. Reactor containment design rules are such that these structures are capable of withstanding loads significantly higher than the design basis values, with the result that, in the event of a severe accident, loss of containment integrity with widespread environmental consequences is unlikely to occur. Nevertheless, in our present state of knowledge, it would not seem justifiable to ignore the failure modes identified about twelve years ago in the RSS VASH 1400:
1) Early or delayed failures (modes alpha, gamma and delta).
2) Basemat melt-through (mode epsilon).
3) Pre-existing openings in the containment wall (mode beta).
4) By-pass hazard.
However, there have been significant developments in the way in which these different hazards are perceived (see paragraph 29).
73. This report is a fairly detailed presentation of the level of knowledge achieved and the surveys and associated methods of assessment enabling the ultimate behaviour of containments to be estimated, according to the type of challenge, including both high internal overpressure and various external hazards. In the light of significant R and D contributions, the conclusions differ little from those reached in the previous Senior Group report. Thus, in regard to internal overpressure effects: "... it would require pressures 1.5 to 2 times higher than the design pressure, and sometimes even more, to cause a containment failure". The lack of precision in this statement is a consequence of the rather wide range of results which are strongly influenced by the specific design of the various systems.
74. The conclusions previously issued on source term aspects, some two or three years ago, both by the Source Term Task Force* and in the Senior Group report, remain valid. However, we would like to stress the importance attached to the consequences of leaks through the containment from: 1) accidents far less improbable than those leading to sudden, extensive release, 2) problems related to the crack trapping factor, 3) the chemical forms of iodine, 4) the various risks of radioactive re-emission within the containment, 5) the part played by auxiliary buildings in the containment function, etc. Priority should be given to pursuance of understanding the phenomenology of the trans fer of radioactive products within the plant boundary.
* Nuclear Reactor Accident Source Terms; Report by an NEA Group of Experts, March 1986.
24 75. With regard to the early failure hazard there is some variation in per ception from one country to another. However, there is general agreement that there is no reason to cast doubt on plant design. It should be noted that, whilst the conditional probability of early rupture is considered adequately low, the occurrence of moderately sized energy release peaks, below a level which would jeopardise containment integrity, is more likely (steam production peak, mini-hydrogen explosion, sudden ejection of molten debris in the event of vessel melt-through, etc.). There is a general consensus on the way in which risks entailed by other failure modes are perceived, which emphasizes the need to prevent or minimise consequences by accident management capabili ty. Either by satisfactory analysis of the design or with the adoption of appropriate reactor control procedures or complementary provisions, the risk of situations likely to lead to early containment ruptcre can be greatly re duced.
76. The accident management aspect is extensively dealt with within the report. However, despite the undeniable potential advantages of this approach, certain problems, such as those listed below, should not be over looked :
- First of all, any method of accident management,with associated means of operator intervention, must in no way detract from the strict observance of design rules nor result in exacerbation of an accident situation.
- The number and range of possible scenarios is considerably greater in the context of severe accident management than for the design basis. The phenomenological sequences can involve significant momentary disturbances of physical parameters (pressure, tempera ture, energy spikes, ultimate destination of fission products, etc.).
- Owing mainly to the complexity of the phenomena involved, accident management efficiency will be based on case by case diagnosis and follow-up ability, together with a fairly rough estimation of the consequences of actions implemented.
However, most countries are convinced that certain crucial parameters can be controlled by accident management, especially by forestalling extreme situa tions involving, in the first place, the ultimate strength of the containment, the assessment of which is subject to high uncertainty.
77. There is thus a marked convergence of views in favour of accident management as a complement or alternative, to be adopted in cases beyond the normal design basis, with a view to preventing or limiting excessive loads on the containment which could ultimately jeopardise its integrity.
78. Present understanding on this subject is adequate for defining the preliminary corrective action guidelines based on the evolution of physical parameters, using a state-by-state or symptom-oriented approach. Such an approach depends not only on the available means of assessment and measure ment, but also, of course, on their continued availability which implies consistently judicious decisions as to how they should be used. Should such initial measures fail, limitation of environmental consequences by accident management would be based on further measures which might differ considerably from one country to another, divergences being mainly due to whether or not a
2' filtered containment system is adopted.The standard characteristicsofrn- tered containment venting and the physical parameters which can be controlled by resorting to them are presented in paragraphs 45 and 67 to 70. It should be noted, in particular, that the use of such strategies enables accident management to be initiate^ more accurately and with greater flexibility, well before the final stage where the question of monitored containment venting would arise. Furthermore the operator would draw confidence during that phase from the knowledge that in any event as a last resort he could implement fil tered venting. However, it must be borne in mind that a filtered-venting system is only to be used when all other means have failed. Priority must always be given to preserving containment integrity.
79. Much of the pooled knowledge derived from extensive R and D actions is useful for both safety assessment and accident management whilst other data are more specifically related to one or other of these aspects. As regards safety assessment, fairly substantial R and D programmes are being implemented, especially vith a view to reducing the far-reachi..g uncertainties which obscure the analysis of severe accidents having the most extreme consequences. It is particularly important to adopt a realistic approach in apprehending phenomena encountered in slowly developing accidents, where there is a real possibility of effective remedial action. In this context, assuming aggravating circum stances or limiting cases could lead to selection of provisions or procedures which would be inappropriate for most of the other cases encountered. Deeper R and D analysis of certain phenomena vould certainly be justified. It is to be noted that the pursuance of such efforts would also be favourable to the improvement and preservation of high level competence, which would be indispensable in the event of a crisis.
80. Current research activities relevant to containment technology manage ment in severe accidents are reasonably extensive and well-served by specialist meetings and conferences and other interchanges between experts. However, a problem remains in harnessing the specialist knowledge and ensuring that its content and significance is fully available to and appreciated by designers and operators who are crlled upon to make wider-ranging decisions.
81. Training of plant peisonnol should be extended to include a broad understanding of severe accident development and a keen awareness of the significance to severe accident management of abnormal states of the plant. The task is large; knowledge, preparation, guides, procedures, training and periodic exercises will be required in order to establish a pervasive "safety culture" at all levels. It will be important to review responsibilities and assign them amongst management and personnel concerned. This burden rests largely on the owners of plants and utilities.
82. The development of '.he nuclear industry has provided many examples of the importance of appreciating and applying the lessons of experience in operation. This is no less true in relation to the management of severe accidents. Accordingly, in analysing unusual operating incidents, possible implications on the ability of the containment and associated equipment to perform its function in a severe accident should be borne in mind. In analys ing accident precursors these aspects should be specially addressed, and the results of analyses exchanged and discussed widely.
26 List of authors
The following members of the Senior Group of Experts on Severe Accidents contributed to this report:
France Mr. Jacques FELCE (CEA) (Vice-Chairman)
Federal Republic Prof. Dr. Enno F. HICKEN (GRS) of Germany
Italy Mr. Carlo ZAFFIRO (ENEA/DISP)
Japan Mr. Toshinori IIJIMA (JINS) Dr. Kunihisa SODA (JAERI)
Sweden Mr. Per BYSTEDT (SKI)
United Kingdom Mr. Peter BARR (UKAEA) Prof. Harry J. TEAGUE (UKAEA) (Chairman)
United States Mr. Daniel F. GIESSING (USDOE) Dr. Denwood F. ROSS (USNRC) Dr. Themis P. SPEIS (USNRC)
Commission of the Mr. Jean-Louis LAMY European Communities
* * *
OECD Nuclear Energy Agency Mr. Jacques ROYEN (Secretary)
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OECD PUBLICATIONS, 2, rm» Andr«-P»»c.l, 75775 PARIS CEDEX 18 - No. 44839 19 i PAINTED IN FRANCE (86 89 08 1) ISBN 92-64-13283-5