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Development of design procedures for fast reactors in the United Kingdom
Int. J. Pres. Ves. & Piping 37 (1989) 99 112
Development of Design Procedures for Fast Reactors in the United Kingdom R. T. Rose National Nuclear Corporation, Warrington Road, Risley, Warrington, Cheshire WA3 6BZ, UK
B. Tomkins Structural Integrity Centre, Risley Nuclear Power Development Laboratories, United Kingdom Atomic Energy Authority--Northern Division, Risley, Warrington, Cheshire WA3 6AT, UK
&
C. H. A. Townley Central Electricity Generating Board, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire, GL13 9PB, UK
ABSTRACT A considerable amount of research has been carried out in the United Kingdom during the past two decades to quantify the factors which control the integrity of structural components. The work which has been aimed at understanding the performance of structures at high temperature is particularly relevant to the Fast Reactor. At the same time, because of the need to demonstrate the tolerance to defects in the low temperature as well as the high temperature components, defect assessment criteria are also of great importance. Emphasis is now being given to the development of design procedures specifically for Fast Reactors, making use of the research so far completed. The United Kingdom proposals are being integrated with those from France, Federal Republic of Germany and Italy as part of the European collaborative venture. The paper outlines the major developments which are currently in hand, and brings up to date the review of United Kingdom activities presented at Tokyo 99 Int. J. Pres. Ves. & Piping 0308-0161/89/$03-50 © 1989 Elsevier Science Publishers Ltd England. Printed in Great Britain
100
R. T. Rose, B. Tomkins, C. H. A. Townley in 1986. An important aspect of UK philosophy is that simplified methods of design should be used wherever possible to avoid time consuming inelastic computations.
INTRODUCTION In the United Kingdom development of design procedures for fast reactor primary and secondary circuits is undertaken by the Structural Integrity Working Group. This is one of a series of working groups which cover all aspects of fast reactor technology and which bring together expertise from the UKAEA, NNC and CEGB. In addition to the development of design procedures, the Structural Integrity Working Group has responsibility for the supporting programme of research. It has close links with the Fast Reactor Materials Working Group and with the groups in CEGB developing assessment procedures for existing nuclear plant. The moves towards joint European funding of future fast reactor development have led to close collaboration between the Structural Integrity Working Group and its counterparts on the Continent. Existing research results are brought together under the auspices of Arbeitsgruppe/Groupe de Travail 9B of the European Research Steering Committee. This body has also drawn up a co-ordinated programme for future research. Design procedures developed in the United Kingdom are now being made available to a committee which has recently been set up by the Design and Construction Companies to consider alterations and additions to the French RCC-MR rules. This activity is seen as a major step forward towards a unified European Fast Reactor Design Code. There is now less emphasis on the development of the independent U K Design Index, described in Ref. 1. In the space available, the present paper can do little more than draw attention to the major developments which are in hand. It brings up to date the review of United Kingdom activities presented to the Conference on Engineering Applications of Creep in Tokyo in 1986.2 Emphasis in the United Kingdom is placed on deriving simplified methods of design, as a means of avoiding, wherever possible, time-consuming inelastic computations. Importance is attached to established methods of defect assessment, involving considerations of fracture and crack propagation. SHAKEDOWN The existence of a shakedown state has provided the basis of design procedures that have been widely accepted. For example it is the origin of the (PL + Pa + Q) rules in Section 1 of the ASME Code. It has also been used to
UK development of design proceduresfor fast reactors
101
derive the design rules for branch connections and other stress concentrating features in codes such as BS5500. Although originally derived for applications below the creep cross over temperature, shakedown concepts have been extended to higher temperatures. 3'4 The method is based on an elastic stress analysis of the component, the only additional requirement being a postulated residual stress system which in itself does not violate yield. It provides bounds on how far the elastically calculated stresses may exceed yield, and ensures considerable relaxation of the over-pessimistic elastic route rules in present fast reactor design codes. Peak stresses do not" have to be included, and can be treated separately. Satisfaction of the shakedown criterion automatically guarantees freedom from ratcheting, and in the case of high temperature components, freedom from excessive creep deformation. Strain ranges in peak stress regions of the component are limited to a little more than their elastic values. A design procedure based on the shakedown method is in the process of being adopted in the United Kingdom for fast reactor purposes. A computer program has been developed to optimise the residual stress system used in the procedure, although often it is possible to do this more simply. An outline of the method and a comparison of the resulting designs with existing elastic route designs is given in Ref. 5. It appears likely that many well designed components will meet the shakedown criterion, and can thus be assessed by simplified methods. A component which does not meet shakedown requirements is not necessarily unacceptable. A full inelastic analysis is, however, likely to be necessary to demonstrate that it is satisfactory.
INTERACTION DIAGRAMS As described previously, the simple Bree diagram provided in Code Case N47 is unsatisfactory for the majority of fast reactor components. 2 A more general approach has been developed by Ponter, who has been partly funded by the Commission of European Communities. 6"7 A series of interaction diagrams have been produced which cater for the situations most often encountered in the fast reactor. The general form of the diagram is shown in Fig. la. It is a relatively simple matter to plot an assessment point on the diagram, knowing the elastically calculated stresses due to external loads and the elastically calculated thermal stresses. If the assessment point falls in the E or S regions, the component will shakedown. As described above, this means that the component will not ratchet and that strains in stress concentration regions are limited to near their elastic values.
102
R. T. Rose, B. Tomkins, C. H. A. Townley
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plasticity 3
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If the point falls within the P region there will be no ratcheting, but large regions of the component will be subjected to plastic cycling. Full consideration will have to be given to the amount of fatigue and creep-fatigue damage. Some form of inelastic analysis will be required to establish strain ranges and stress levels, although wherever possible this will be based on elastic calculations, coupled with strain range enhancement factors. If the point falls within the R region ratcheting will occur as well as plastic cycling. The present view is that a component which falls within the R region is unlikely to be satisfactory. This is reinforced by the finding that, in some situations, an unacceptably large amount of ratcheting occurs as soon as the P - R boundary or S-R boundary is crossed. The present phase of development is to produce a simple design procedure which will include a compendium of interaction diagrams covering the cases of importance. A computer program is available which can be used to deal
UK development of design proceduresfor fast reactors
103
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with situations which are not included in the compendium. Experimental validation is in hand at Leicester University, UK, and using the DEFT rig in UKAEA. A typical diagram such as will be provided in the compendium is shown in Fig. lb. This is for a tube subjected to a moving temperature front. The differences between this diagram and the simple Bree diagram of Code Case N47 are immediately apparent, particularly at low values of trp/try which are most relevant to fast reactor components. It is interesting to note that there is no P region for this situation. The component either shakes down or it ratchets. T H E R M A L SHOCK AND T H E R M A L STRIPING Thermal shock and thermal striping are key factors which have to be taken into account in fast reactor structural design. Thermal shock can occur
104
R. T. Rose, B. Tornkins, C. H. A. Townley
during a reactor trip and other disturbances to steady operation. Thermal striping occurs wherever two streams of sodium at different temperatures mix together. Components subjected to thermal striping are usually subjected also to thermal shock. The reverse is not necessarily true. The starting point for the design of a component is a full thermal hydraulics study. While this subject is outside the scope of the present paper, it is important to note that such a study must define the temperature variations in the fluid and the associated heat flows into the component. So far attention has been focused on those situations, believed to be the majority, where local effects predominate. It is, however, important to identify and take into consideration any long range thermal stresses which may be induced. In the absence of long range effects, stresses arising from thermal shock and thermal striping will largely be localised on or near the surface. Shakedown will be achieved and damage in the bulk of the material will be relatively unimportant. The United Kingdom approach is therefore to treat the problem as one of crack initiation and crack propagation. An interim design procedure has been adopted for components at core outlet temperature subjected to both shock and striping. This is based on the observation that cracks are initiated by the creep damage which occurs during stress relaxation following thermal shocks. Safe working life is then determined by the crack propagation which takes place during thermal striping. A flow diagram for the procedure is shown in Fig. 2. Applied to a typical above core structure, the crack initiation-crack propagation procedure provides a limit on allowable temperature range about twice as great as that permitted by present design rules. Further alleviations are likely to be achieved by refining crack propagation calculations and by taking into account lack of coherence of the thermal boundary conditions. The ongoing development of the C L O U D B U R S T computer program will provide a better understanding of the factors /ti[Evaluate creep damage Correction for reduced / relaxation /
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UK development of design procedures for.fast reactors
105
controlling crack propagation, including lack of coherence. Experimental confirmation of the procedure will come from tests on the S U P E R S O M I T E facility, which will shortly supplement the older SOMITE rig. As an interim measure, creep and fatigue damage estimation procedures similar to those in Code Case N47 and R C C - M R have been adopted. However, the view in the United Kingdom is that these do not provide a true representation of the mechanisms of failure and do not emphasise the important role of high ductility in enhancing resistance to failure. Progress towards a creep--fatigue damage assessment based on ductility exhaustion has been described previously. 8 Design procedures based on this concept are expected to be available in about 12 months. Experimental backing comes from thermal shock rigs operated by the UKAEA, N N C and CEGB. F R A C T U R E A N D LEAK BEFORE BREAK A subject which is not treated in existing high temperature design codes and one which is of considerable importance is the assessment of crack-like defects. At the design stage it is necessary to show that components are tolerant to the presence of hypothetical defects of a size which might escape detection during manufacture. It is also necessary to show that any actual defects which are discovered during manufacture or during operation do not compromise safety and reliability. There are two stages in the assessment of defects. Firstly the critical defect size for fast fracture must be estimated. Then the way in which a hypothetical or actual defect can grow towards this critical size must be examined. Leak before break arguments are invoked to give additional assurance of safety. Because of the high ductility of stainless steels, it is obvious that linear elastic fracture mechanics is inappropriate to determine critical crack sizes. The CEGB R6 procedure, which allows for post yield effects, has therefore been adopted, with certain modifications. 9 Further details are given in the Tokyo Conference paper, previously mentioned. 2 The alternative would be to carry out a full inelastic finite element analysis to calculate J directly, but this is seen as too expensive and time-consuming, except in special circumstances. Crack growth can occur, at the lower temperatures, either by fatigue or by stable tearing during abnormal events. For cracks with plane fronts fatigue growth can adequately be described by laws of the Paris type, which relate rate to AK. Stable tearing is estimated from tearing resistance data. At the higher temperatures creep crack growth and creep-fatigue crack growth are the important issues. Considerable progress has been made in understanding the factors which
06
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UK development of design proceduresfor fast reactors
107
control these phenomena. Assessment procedures have been developed, using simplified methods of analysis, for application to high temperature components of gas cooled reactors and fossil fired power plant. 10,11 These are now being considered for fast reactor application. A flow chart for assessment of creep crack propagation is given in Fig. 3. Leak before break relies on demonstrating that an initial defect will propagate in such a way that detectable leakage will occur well before the crack reaches the critical crack size for fast fracture. Data are not needed on crack propagation rates nor on the number of loading cycles which are applied to components. What is important is a reasonably accurate knowledge of the way in which the shape of the crack changes as it propagates. For fatigue crack propagation at the lower temperatures, methods of predicting changes in crack shape are available for components subjected to membrane and membrane plus bending stresses. The position is less satisfactory for those components which have high surface thermal stresses, but here the primary safety case is likely to be based on limits on crack propagation rather than leak before break. For creep and creep-fatigue crack propagation, the present intention is to adapt the methods which are used to predict changes in shape of fatigue cracks. Changes in crack shape due to stable tearing have so far not been treated theoretically. Tests are currently in hand on the U K A E A 20MN wide plate machine and elsewhere to validate leak before break procedures at the lower temperatures, and to explore changes in crack shape during stable tearing experimentally. Similar work at high temperature is likely to prove expensive, and is not included in the programme for the immediate future. The Structural Features Test Facility, which is under construction by the UKAEA, will be used to validate fracture and crack propagation predictions at near full size. Intended initially for operation at moderate temperatures, it will first be used to examine components such as the core support, the triple point and the transition joint between the main tank and the roof. Investigations to validate leak before break in the secondary pipework are already in hand in the UKAEA, using a smaller hydraulic rig. Fatigue cracks are grown in a tee-connection, and the results compared with theoretical predictions. DYNAMIC LOADING There are two types of dynamic loading which have to be taken into account: vibrations which occur in service, often during off-normal operation, and those associated with events such as missile impact and earthquakes.
108
R. T. Rose, B. Tomkins, C. H. A. Townley
In the former category, issues are often design specific, and require individual investigation. The development of general design procedures is considered inappropriate, although limits on allowable stresses will be the same as for other sorts of loading. The second category--improbable events which nevertheless have to be taken into account during design--form the subject of a co-ordinated nuclear industries' research programme. Preliminary procedures have been drafted for the impact of missiles on concrete structures. Guidelines on pipewhip are expected shortly. Design for seismic loading is of particular interest in the fast reactor context. Thickness of primary circuit components such as the main vessel, the inner tank and the above core structure can be determined by the need to resist earthquakes. To make such structures thicker than required to meet service conditions not only leads to unnecessary expense; it may also lead to premature failure in service because thermal stresses will be increased by the increased thickness. For these reasons there is a good case for isolating the reactor building on
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UK development of design procedures for fast reactors
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aseismic mountings. Theoretical investigations have shown that this is a promising solution and attention is now being directed towards the development of suitable devices. A typical result is shown in Fig. 4, which emphasises the differences in accelerations at various heights in the reactor building, with and without isolation. In parallel, work is being directed towards the elimination of snubbers from pipework systems. Snubbers are known to require extensive maintenance, which could lead to long plant shutdowns and, in some cases, to increased radiation exposure of staff. Seizure during normal operation can produce excessively high stresses in components that need to be free to expand and contract. Secondary pipework has been selected for the initial studies and theoretical investigations have been carried out on a typical system operating at 510°C. An optimised mix of constant load and rigid supports can be provided so that code limits on seismic stresses and in-service stresses are met with only a minimal snubber requirement. The initial design included as many as 15 snubbers; the final design only two. It should also be noted that the code limits for seismic stresses are themselves pessimistic. Tests on pipe sections and bends, under simulated earthquake loading, have shown that an elastic dynamic analysis combined with present code stress limits grossly underestimates the true safety margins. In reality, stresses and strains are limited because damping increases significantly as the most highly stressed regions of the pipe become plastic. 12 Figure 5 provides a plot of experimental results and clearly demonstrates the self-limiting nature of the acceleration at the centre of the piping span as the acceleration of the supports is increased.
110
R. T. Rose, B. Tomkins, C. H. A. Townley
M A T E R I A L S DATA A N D WELDS The design procedures which are being developed for the fast reactor require codified materials data which are far more extensive than normally found in traditional codes and standards. This applies quite generally, and not just to the United Kingdom procedures, and arises because of the nature of the loadings which have to be considered in the design of fast reactor components. At the higher temperatures, cyclic thermal stressing is dominant. Conventional creep deformation data and creep rupture data derived from constant load tests are largely irrelevant. Cyclic deformation data are needed to derive constitutive equations and cyclic failure data are needed to establish permissible stress and strain ranges. Since ductility is often more important than strength in determining the life of a component, data are needed on failure strains. With the increasing emphasis on crack initiation and propagation in the design of critical components, data are now needed on propagation rates under fatigue, creep and creep-fatigue conditions. Data on thresholds for non-propagating fatigue cracks and on the incubation time for creep cracks are also needed. Consideration of events such as missile impacts and sodium-water reactions introduces a requirement for data on materials under high rates of strain. Deformation, failure stress and failure strain are all important. There is a need throughout to take into account changes in materials properties which occur during the life of the plant. This is particularly true of data used to establish limits on permitted stresses and strains and those associated with the assessment of defects. Embrittlement due to radiation and to thermal ageing must be taken into account. Environmental effects leading to stress corrosion embrittlement may also need to be considered. Finally there is the all-important question of welds. Since experience shows that welds are usually the life limiting feature of high temperature plant it is questionable whether the present emphasis on parent material properties is correct. What is important is the performance of the complete weldment, not the weld metal properties per se, and this is strongly influenced by the method of welding and by details of the welding process. The United Kingdom view is that much more test work is needed on complete weldments, under the types of loading which occur in the fast reactor, before design procedures can be fully established. The present belief is that designers will estimate stresses and strains in a component as if it were homogeneous, and then apply strength reduction factors or life reduction factors to allow for the presence of the weld. In the case of defect assessment they would use crack propagation data appropriate to the weldment, after allowing for any long term degradation.
UK development of design proceduresfor fast reactors
111
So far United Kingdom work on weldments has mainly been directed towards high cycle fatigue. Attention is now being turned towards tests on weldments under conditions which simulate thermal shock and combined thermal shock and thermal striping. This is being backed up by theoretical investigations aimed at understanding the factors which control weld performance. ACKNOWLEDGEMENT This paper is published by permission of the Central Electricity Generating Board, the National Nuclear Corporation and the United Kingdom Atomic Energy Authority. The authors wish to thank all their colleagues who have assisted in the preparation of this paper, and have contributed much of the detailed information. REFERENCES 1. Rose, R. T., Overview of UK position on design codes for LMFBR. 2nd International Seminar on Standards and Structural Analysis in Elevated Temperature Applications for Reactor Technology, ENEA, Venice, October 1986. 2. Rose, R. T., Tomkins, B. & Townley, C. H. A., The development of design criteria for fast reactors in the UK. International Conference on Creep, Tokyo, 1986. 3. Goodall, I. W., Leckie, F. J., Ponter, A. R. S. & Townley, C. H. A., Development of high temperature design methods based on reference stress., J. Engng Mater. Technol. 101 (1979) 349-55. 4. Ainsworth, R. A., Goodall, I. W. & Waiters, D. J., Cyclic loading in the creep range. International Conference on Engineering Aspects of Creep, l.Mech.E., Sheffield, 1980. 5. Rose, R. T., Elastic route design based on creep-modified shakedown. 2nd International Seminar on Standards and Structural Analysis in Elevated Temperature Applications for Reactor Technology, ENEA, Venice, October 1986. 6. Ponter, A. R. S., Shakedown and ratchetting below the creep range. Commission of European Communities Report EUR 8702, 1983. 7. Ponter, A. R. S., Ratchetting in the creep range. Commission of European Communities Report EUR 9876, 1985. 8. Bestwick, R. D. & Clayton, A. M., Design methodology for creep fatigue assessment using creep ductility criteria. 5th International Symposium on Inelastic Analysis in Life Prediction in High Temperature Environments, Paris, August 1985. 9. Milne, I., Dowling, A. R., Ainsworth, R. A. & Stewart, A. T., Assessment of the integrity of structures containing defects. CEGB Assessment Procedure R6, 1987.
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R. T. Rose, B. Tornkins, C. H. A. Townley
10. Ainsworth, R. A. et al., CEGB assessment procedure for defects in plant operating in the creep range. Fatigue and Fracture Engineering Materials and Structures, 10 (1987) 115-27. 11. Ainsworth, R. A. & Goodall, I. W., Development of procedures for the assessment of defects at elevated temperature. 2nd International Seminar on Design Codes and Structural Mechanics, Lausanne, 1987. 12. Beaney, E. M., The response of pipes to seismic loading. 9th International Conference on Structural Mechanics in Reactor Technology, Lausanne, 1987.
Development of Design Procedures for Fast Reactors in the United Kingdom R. T. Rose National Nuclear Corporation, Warrington Road, Risley, Warrington, Cheshire WA3 6BZ, UK
B. Tomkins Structural Integrity Centre, Risley Nuclear Power Development Laboratories, United Kingdom Atomic Energy Authority--Northern Division, Risley, Warrington, Cheshire WA3 6AT, UK
&
C. H. A. Townley Central Electricity Generating Board, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire, GL13 9PB, UK
ABSTRACT A considerable amount of research has been carried out in the United Kingdom during the past two decades to quantify the factors which control the integrity of structural components. The work which has been aimed at understanding the performance of structures at high temperature is particularly relevant to the Fast Reactor. At the same time, because of the need to demonstrate the tolerance to defects in the low temperature as well as the high temperature components, defect assessment criteria are also of great importance. Emphasis is now being given to the development of design procedures specifically for Fast Reactors, making use of the research so far completed. The United Kingdom proposals are being integrated with those from France, Federal Republic of Germany and Italy as part of the European collaborative venture. The paper outlines the major developments which are currently in hand, and brings up to date the review of United Kingdom activities presented at Tokyo 99 Int. J. Pres. Ves. & Piping 0308-0161/89/$03-50 © 1989 Elsevier Science Publishers Ltd England. Printed in Great Britain
100
R. T. Rose, B. Tomkins, C. H. A. Townley in 1986. An important aspect of UK philosophy is that simplified methods of design should be used wherever possible to avoid time consuming inelastic computations.
INTRODUCTION In the United Kingdom development of design procedures for fast reactor primary and secondary circuits is undertaken by the Structural Integrity Working Group. This is one of a series of working groups which cover all aspects of fast reactor technology and which bring together expertise from the UKAEA, NNC and CEGB. In addition to the development of design procedures, the Structural Integrity Working Group has responsibility for the supporting programme of research. It has close links with the Fast Reactor Materials Working Group and with the groups in CEGB developing assessment procedures for existing nuclear plant. The moves towards joint European funding of future fast reactor development have led to close collaboration between the Structural Integrity Working Group and its counterparts on the Continent. Existing research results are brought together under the auspices of Arbeitsgruppe/Groupe de Travail 9B of the European Research Steering Committee. This body has also drawn up a co-ordinated programme for future research. Design procedures developed in the United Kingdom are now being made available to a committee which has recently been set up by the Design and Construction Companies to consider alterations and additions to the French RCC-MR rules. This activity is seen as a major step forward towards a unified European Fast Reactor Design Code. There is now less emphasis on the development of the independent U K Design Index, described in Ref. 1. In the space available, the present paper can do little more than draw attention to the major developments which are in hand. It brings up to date the review of United Kingdom activities presented to the Conference on Engineering Applications of Creep in Tokyo in 1986.2 Emphasis in the United Kingdom is placed on deriving simplified methods of design, as a means of avoiding, wherever possible, time-consuming inelastic computations. Importance is attached to established methods of defect assessment, involving considerations of fracture and crack propagation. SHAKEDOWN The existence of a shakedown state has provided the basis of design procedures that have been widely accepted. For example it is the origin of the (PL + Pa + Q) rules in Section 1 of the ASME Code. It has also been used to
UK development of design proceduresfor fast reactors
101
derive the design rules for branch connections and other stress concentrating features in codes such as BS5500. Although originally derived for applications below the creep cross over temperature, shakedown concepts have been extended to higher temperatures. 3'4 The method is based on an elastic stress analysis of the component, the only additional requirement being a postulated residual stress system which in itself does not violate yield. It provides bounds on how far the elastically calculated stresses may exceed yield, and ensures considerable relaxation of the over-pessimistic elastic route rules in present fast reactor design codes. Peak stresses do not" have to be included, and can be treated separately. Satisfaction of the shakedown criterion automatically guarantees freedom from ratcheting, and in the case of high temperature components, freedom from excessive creep deformation. Strain ranges in peak stress regions of the component are limited to a little more than their elastic values. A design procedure based on the shakedown method is in the process of being adopted in the United Kingdom for fast reactor purposes. A computer program has been developed to optimise the residual stress system used in the procedure, although often it is possible to do this more simply. An outline of the method and a comparison of the resulting designs with existing elastic route designs is given in Ref. 5. It appears likely that many well designed components will meet the shakedown criterion, and can thus be assessed by simplified methods. A component which does not meet shakedown requirements is not necessarily unacceptable. A full inelastic analysis is, however, likely to be necessary to demonstrate that it is satisfactory.
INTERACTION DIAGRAMS As described previously, the simple Bree diagram provided in Code Case N47 is unsatisfactory for the majority of fast reactor components. 2 A more general approach has been developed by Ponter, who has been partly funded by the Commission of European Communities. 6"7 A series of interaction diagrams have been produced which cater for the situations most often encountered in the fast reactor. The general form of the diagram is shown in Fig. la. It is a relatively simple matter to plot an assessment point on the diagram, knowing the elastically calculated stresses due to external loads and the elastically calculated thermal stresses. If the assessment point falls in the E or S regions, the component will shakedown. As described above, this means that the component will not ratchet and that strains in stress concentration regions are limited to near their elastic values.
102
R. T. Rose, B. Tomkins, C. H. A. Townley
Tensile Ltchetin R1
R2
plasticity 3
PI
¢I w II
2
~hokedown $1
0 V Fully elasticl behaviour~ E -1
-0.5
Fig. la.
o
P/A a-y
o15
1
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If the point falls within the P region there will be no ratcheting, but large regions of the component will be subjected to plastic cycling. Full consideration will have to be given to the amount of fatigue and creep-fatigue damage. Some form of inelastic analysis will be required to establish strain ranges and stress levels, although wherever possible this will be based on elastic calculations, coupled with strain range enhancement factors. If the point falls within the R region ratcheting will occur as well as plastic cycling. The present view is that a component which falls within the R region is unlikely to be satisfactory. This is reinforced by the finding that, in some situations, an unacceptably large amount of ratcheting occurs as soon as the P - R boundary or S-R boundary is crossed. The present phase of development is to produce a simple design procedure which will include a compendium of interaction diagrams covering the cases of importance. A computer program is available which can be used to deal
UK development of design proceduresfor fast reactors
103
~J~y R Rotcheting
$
Futly elastic behaviour " ~
0 Fig. lb.
0"iVO,,y
I
Interaction diagram for tube subjected to oscillating temperature front.
with situations which are not included in the compendium. Experimental validation is in hand at Leicester University, UK, and using the DEFT rig in UKAEA. A typical diagram such as will be provided in the compendium is shown in Fig. lb. This is for a tube subjected to a moving temperature front. The differences between this diagram and the simple Bree diagram of Code Case N47 are immediately apparent, particularly at low values of trp/try which are most relevant to fast reactor components. It is interesting to note that there is no P region for this situation. The component either shakes down or it ratchets. T H E R M A L SHOCK AND T H E R M A L STRIPING Thermal shock and thermal striping are key factors which have to be taken into account in fast reactor structural design. Thermal shock can occur
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R. T. Rose, B. Tornkins, C. H. A. Townley
during a reactor trip and other disturbances to steady operation. Thermal striping occurs wherever two streams of sodium at different temperatures mix together. Components subjected to thermal striping are usually subjected also to thermal shock. The reverse is not necessarily true. The starting point for the design of a component is a full thermal hydraulics study. While this subject is outside the scope of the present paper, it is important to note that such a study must define the temperature variations in the fluid and the associated heat flows into the component. So far attention has been focused on those situations, believed to be the majority, where local effects predominate. It is, however, important to identify and take into consideration any long range thermal stresses which may be induced. In the absence of long range effects, stresses arising from thermal shock and thermal striping will largely be localised on or near the surface. Shakedown will be achieved and damage in the bulk of the material will be relatively unimportant. The United Kingdom approach is therefore to treat the problem as one of crack initiation and crack propagation. An interim design procedure has been adopted for components at core outlet temperature subjected to both shock and striping. This is based on the observation that cracks are initiated by the creep damage which occurs during stress relaxation following thermal shocks. Safe working life is then determined by the crack propagation which takes place during thermal striping. A flow diagram for the procedure is shown in Fig. 2. Applied to a typical above core structure, the crack initiation-crack propagation procedure provides a limit on allowable temperature range about twice as great as that permitted by present design rules. Further alleviations are likely to be achieved by refining crack propagation calculations and by taking into account lack of coherence of the thermal boundary conditions. The ongoing development of the C L O U D B U R S T computer program will provide a better understanding of the factors /ti[Evaluate creep damage Correction for reduced / relaxation /
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UK development of design procedures for.fast reactors
105
controlling crack propagation, including lack of coherence. Experimental confirmation of the procedure will come from tests on the S U P E R S O M I T E facility, which will shortly supplement the older SOMITE rig. As an interim measure, creep and fatigue damage estimation procedures similar to those in Code Case N47 and R C C - M R have been adopted. However, the view in the United Kingdom is that these do not provide a true representation of the mechanisms of failure and do not emphasise the important role of high ductility in enhancing resistance to failure. Progress towards a creep--fatigue damage assessment based on ductility exhaustion has been described previously. 8 Design procedures based on this concept are expected to be available in about 12 months. Experimental backing comes from thermal shock rigs operated by the UKAEA, N N C and CEGB. F R A C T U R E A N D LEAK BEFORE BREAK A subject which is not treated in existing high temperature design codes and one which is of considerable importance is the assessment of crack-like defects. At the design stage it is necessary to show that components are tolerant to the presence of hypothetical defects of a size which might escape detection during manufacture. It is also necessary to show that any actual defects which are discovered during manufacture or during operation do not compromise safety and reliability. There are two stages in the assessment of defects. Firstly the critical defect size for fast fracture must be estimated. Then the way in which a hypothetical or actual defect can grow towards this critical size must be examined. Leak before break arguments are invoked to give additional assurance of safety. Because of the high ductility of stainless steels, it is obvious that linear elastic fracture mechanics is inappropriate to determine critical crack sizes. The CEGB R6 procedure, which allows for post yield effects, has therefore been adopted, with certain modifications. 9 Further details are given in the Tokyo Conference paper, previously mentioned. 2 The alternative would be to carry out a full inelastic finite element analysis to calculate J directly, but this is seen as too expensive and time-consuming, except in special circumstances. Crack growth can occur, at the lower temperatures, either by fatigue or by stable tearing during abnormal events. For cracks with plane fronts fatigue growth can adequately be described by laws of the Paris type, which relate rate to AK. Stable tearing is estimated from tearing resistance data. At the higher temperatures creep crack growth and creep-fatigue crack growth are the important issues. Considerable progress has been made in understanding the factors which
06
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UK development of design proceduresfor fast reactors
107
control these phenomena. Assessment procedures have been developed, using simplified methods of analysis, for application to high temperature components of gas cooled reactors and fossil fired power plant. 10,11 These are now being considered for fast reactor application. A flow chart for assessment of creep crack propagation is given in Fig. 3. Leak before break relies on demonstrating that an initial defect will propagate in such a way that detectable leakage will occur well before the crack reaches the critical crack size for fast fracture. Data are not needed on crack propagation rates nor on the number of loading cycles which are applied to components. What is important is a reasonably accurate knowledge of the way in which the shape of the crack changes as it propagates. For fatigue crack propagation at the lower temperatures, methods of predicting changes in crack shape are available for components subjected to membrane and membrane plus bending stresses. The position is less satisfactory for those components which have high surface thermal stresses, but here the primary safety case is likely to be based on limits on crack propagation rather than leak before break. For creep and creep-fatigue crack propagation, the present intention is to adapt the methods which are used to predict changes in shape of fatigue cracks. Changes in crack shape due to stable tearing have so far not been treated theoretically. Tests are currently in hand on the U K A E A 20MN wide plate machine and elsewhere to validate leak before break procedures at the lower temperatures, and to explore changes in crack shape during stable tearing experimentally. Similar work at high temperature is likely to prove expensive, and is not included in the programme for the immediate future. The Structural Features Test Facility, which is under construction by the UKAEA, will be used to validate fracture and crack propagation predictions at near full size. Intended initially for operation at moderate temperatures, it will first be used to examine components such as the core support, the triple point and the transition joint between the main tank and the roof. Investigations to validate leak before break in the secondary pipework are already in hand in the UKAEA, using a smaller hydraulic rig. Fatigue cracks are grown in a tee-connection, and the results compared with theoretical predictions. DYNAMIC LOADING There are two types of dynamic loading which have to be taken into account: vibrations which occur in service, often during off-normal operation, and those associated with events such as missile impact and earthquakes.
108
R. T. Rose, B. Tomkins, C. H. A. Townley
In the former category, issues are often design specific, and require individual investigation. The development of general design procedures is considered inappropriate, although limits on allowable stresses will be the same as for other sorts of loading. The second category--improbable events which nevertheless have to be taken into account during design--form the subject of a co-ordinated nuclear industries' research programme. Preliminary procedures have been drafted for the impact of missiles on concrete structures. Guidelines on pipewhip are expected shortly. Design for seismic loading is of particular interest in the fast reactor context. Thickness of primary circuit components such as the main vessel, the inner tank and the above core structure can be determined by the need to resist earthquakes. To make such structures thicker than required to meet service conditions not only leads to unnecessary expense; it may also lead to premature failure in service because thermal stresses will be increased by the increased thickness. For these reasons there is a good case for isolating the reactor building on
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UK development of design procedures for fast reactors
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aseismic mountings. Theoretical investigations have shown that this is a promising solution and attention is now being directed towards the development of suitable devices. A typical result is shown in Fig. 4, which emphasises the differences in accelerations at various heights in the reactor building, with and without isolation. In parallel, work is being directed towards the elimination of snubbers from pipework systems. Snubbers are known to require extensive maintenance, which could lead to long plant shutdowns and, in some cases, to increased radiation exposure of staff. Seizure during normal operation can produce excessively high stresses in components that need to be free to expand and contract. Secondary pipework has been selected for the initial studies and theoretical investigations have been carried out on a typical system operating at 510°C. An optimised mix of constant load and rigid supports can be provided so that code limits on seismic stresses and in-service stresses are met with only a minimal snubber requirement. The initial design included as many as 15 snubbers; the final design only two. It should also be noted that the code limits for seismic stresses are themselves pessimistic. Tests on pipe sections and bends, under simulated earthquake loading, have shown that an elastic dynamic analysis combined with present code stress limits grossly underestimates the true safety margins. In reality, stresses and strains are limited because damping increases significantly as the most highly stressed regions of the pipe become plastic. 12 Figure 5 provides a plot of experimental results and clearly demonstrates the self-limiting nature of the acceleration at the centre of the piping span as the acceleration of the supports is increased.
110
R. T. Rose, B. Tomkins, C. H. A. Townley
M A T E R I A L S DATA A N D WELDS The design procedures which are being developed for the fast reactor require codified materials data which are far more extensive than normally found in traditional codes and standards. This applies quite generally, and not just to the United Kingdom procedures, and arises because of the nature of the loadings which have to be considered in the design of fast reactor components. At the higher temperatures, cyclic thermal stressing is dominant. Conventional creep deformation data and creep rupture data derived from constant load tests are largely irrelevant. Cyclic deformation data are needed to derive constitutive equations and cyclic failure data are needed to establish permissible stress and strain ranges. Since ductility is often more important than strength in determining the life of a component, data are needed on failure strains. With the increasing emphasis on crack initiation and propagation in the design of critical components, data are now needed on propagation rates under fatigue, creep and creep-fatigue conditions. Data on thresholds for non-propagating fatigue cracks and on the incubation time for creep cracks are also needed. Consideration of events such as missile impacts and sodium-water reactions introduces a requirement for data on materials under high rates of strain. Deformation, failure stress and failure strain are all important. There is a need throughout to take into account changes in materials properties which occur during the life of the plant. This is particularly true of data used to establish limits on permitted stresses and strains and those associated with the assessment of defects. Embrittlement due to radiation and to thermal ageing must be taken into account. Environmental effects leading to stress corrosion embrittlement may also need to be considered. Finally there is the all-important question of welds. Since experience shows that welds are usually the life limiting feature of high temperature plant it is questionable whether the present emphasis on parent material properties is correct. What is important is the performance of the complete weldment, not the weld metal properties per se, and this is strongly influenced by the method of welding and by details of the welding process. The United Kingdom view is that much more test work is needed on complete weldments, under the types of loading which occur in the fast reactor, before design procedures can be fully established. The present belief is that designers will estimate stresses and strains in a component as if it were homogeneous, and then apply strength reduction factors or life reduction factors to allow for the presence of the weld. In the case of defect assessment they would use crack propagation data appropriate to the weldment, after allowing for any long term degradation.
UK development of design proceduresfor fast reactors
111
So far United Kingdom work on weldments has mainly been directed towards high cycle fatigue. Attention is now being turned towards tests on weldments under conditions which simulate thermal shock and combined thermal shock and thermal striping. This is being backed up by theoretical investigations aimed at understanding the factors which control weld performance. ACKNOWLEDGEMENT This paper is published by permission of the Central Electricity Generating Board, the National Nuclear Corporation and the United Kingdom Atomic Energy Authority. The authors wish to thank all their colleagues who have assisted in the preparation of this paper, and have contributed much of the detailed information. REFERENCES 1. Rose, R. T., Overview of UK position on design codes for LMFBR. 2nd International Seminar on Standards and Structural Analysis in Elevated Temperature Applications for Reactor Technology, ENEA, Venice, October 1986. 2. Rose, R. T., Tomkins, B. & Townley, C. H. A., The development of design criteria for fast reactors in the UK. International Conference on Creep, Tokyo, 1986. 3. Goodall, I. W., Leckie, F. J., Ponter, A. R. S. & Townley, C. H. A., Development of high temperature design methods based on reference stress., J. Engng Mater. Technol. 101 (1979) 349-55. 4. Ainsworth, R. A., Goodall, I. W. & Waiters, D. J., Cyclic loading in the creep range. International Conference on Engineering Aspects of Creep, l.Mech.E., Sheffield, 1980. 5. Rose, R. T., Elastic route design based on creep-modified shakedown. 2nd International Seminar on Standards and Structural Analysis in Elevated Temperature Applications for Reactor Technology, ENEA, Venice, October 1986. 6. Ponter, A. R. S., Shakedown and ratchetting below the creep range. Commission of European Communities Report EUR 8702, 1983. 7. Ponter, A. R. S., Ratchetting in the creep range. Commission of European Communities Report EUR 9876, 1985. 8. Bestwick, R. D. & Clayton, A. M., Design methodology for creep fatigue assessment using creep ductility criteria. 5th International Symposium on Inelastic Analysis in Life Prediction in High Temperature Environments, Paris, August 1985. 9. Milne, I., Dowling, A. R., Ainsworth, R. A. & Stewart, A. T., Assessment of the integrity of structures containing defects. CEGB Assessment Procedure R6, 1987.
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10. Ainsworth, R. A. et al., CEGB assessment procedure for defects in plant operating in the creep range. Fatigue and Fracture Engineering Materials and Structures, 10 (1987) 115-27. 11. Ainsworth, R. A. & Goodall, I. W., Development of procedures for the assessment of defects at elevated temperature. 2nd International Seminar on Design Codes and Structural Mechanics, Lausanne, 1987. 12. Beaney, E. M., The response of pipes to seismic loading. 9th International Conference on Structural Mechanics in Reactor Technology, Lausanne, 1987.