Comparison of the partner s approaches for physical phenomena assessment in level 2 PSA

Transcription

1 1/11 Comparison of the partner s approaches for physical phenomena assessment in level 2 PSA J. Eyink, Framatome ANP; H. Löffler, GRS Cologne Abstract Aim of the task 5.1 of the SARNET PSA2 work package is to compare methodologies used by 11 partners for the development of level 2 PSA and to identify improvement needs. One of the subtasks deals with physical phenomena considered in the accident progression event tree (APET). This paper presents the findings of the comparison for the following choice of phenomena: Induced RCS failure and bypass. Vessel failure mode and consequences (e.g. direct containment heating). Hydrogen combustion. Debris coolability during ex-vessel phase. The approaches chosen by the partners will be explained and compared, and needs for the improvement of methods will be identified. 1. INTRODUCTION This presentation is based on the SARNET Working Package 5.1 (WP5.1). The first step of this WP5.1 concerns the description of partner s methodology based on a questionnaire. The second step is the comparison of methodologies used for level 2 PSA development. Methodologies comparison is accompanied by a first review of improvement needs for the different issues. The scope of the comparison is: - level 1 and level 2 PSA interface, - tools used for the study, - accident progression event tree (APET), including systems behaviour, containment leakages and human reliability, - physical phenomena, - releases assessment, - uncertainties assessment general methods, - general results as an input for R & D priorities. Within the topic physical phenomena the following issues have been addressed: 1. Reactor coolant system (RCS) depressurisation and coolability, 2. In vessel core reflooding, 3. Induced RCS failure and bypass, 4. In-vessel fuel coolant interaction, 5. Vessel failure mode and consequences (e.g. direct containment heating), 6. Ex-vessel fuel coolant interaction, 7. Hydrogen combustion, 8. Containment pressurisation and failure, 9. Debris coolability in ex-vessel phase, 10. Other phenomena or effects.

2 2/11 Since 11 partner organisations are involved, and each of the issues above comprises a wide field of aspects, only a small part, related to physical phenomena 3, 5, 7, and 9 will be addressed here. The respective approaches chosen by the partners are explained and compared, and needs for the improvement of methods are identified. The complete comparison can be found in report [1]. The comparison is partly difficult because phenomena for LWR and RBMK are very different. Even the comparison of western LWR and VVER is difficult because of different containment failure modes. Furthermore, the participants often did not identify which relevance and consequences of particular phenomena they are considering. As an example, invessel melt water interaction could be assessed with the classical alpha-mode failure in mind, or in view of melt dispersal within the RCS with related potential thermal loads. These restrictions should be kept in mind when looking at the following comparisons. 2. INDUCED RCS FAILURE AND BYPASS High pressure core melts pose significant risk due to the containment threat when the reactor vessel bottom fails, or due to an induced containment bypass via the steam generator. Manual or automatic depressurisation is foreseen in many plants as an emergency measure against such events. But PSA results show that even then a significant fraction of core melt scenarios may have high pressure. Therefore this issue is important - as a passive means to depressurize the RCS and to avoid the consequences if high pressure reactor pressure vessel (RPV) failure. - to quantify containment bypass. The following table is taken from the SARNET WP5.1 report [1]. Organization Treatment of the phenomenon in the APET AVN Doel 1/2: Induced failure of hot leg, surge line and steam generator tube rupture (SGTR) are considered in the APET Induced SGTR during in-vessel phase Tihange 1: RCS pressure evolution (8 top events) and induced SGTR (3 top events) considered in sub-trees EDF Induced primary leak and SGTR are considered in the APET and quantified by expert judgement FRAM- Based on thermo hydraulic calculations with MAAP and structure mechanical ATOME investigations for induced RCS (no bypass) failure and SGTR (bypass) Interfacing system LOCA with direct release to the environment, with retention in the safeguard building and ground release or partly filtered stack release depending on the size of the break and its location. GRS Konvoi-PWR: Detailed structural analysis of RCS loop shows that hot leg has high failure probability under high pressure core melt conditions. Probability for stuck open safety valves is significant. Probability for induced SGTR is negligible because other failure modes are dominant. 1300MWe-BWR: RCS loop failures are unlikely. Probability for stuck open safety valves is significant IRSN Induced failure of hot leg, surge line and SGTR are considered in the APET. Specific physical model for induced SGTR. Physical analysis done using several associated and complementary means:

3 3/11 LEI NNC SWED- POWER TUS UJV VEIKI - MELCOR then ICARE/CATHARE calculations for relevant scenarios - Detailed thermal hydraulics analysis with CFD codes still ongoing (using CFX and CEA TRIO U code) to assess (as input for the ICARE/CATHARE calculations) the conditions of mixing in the steam generator water box, the steam natural circulation in the hot legs, - Thermo mechanical experiments of hot legs materials (including welding zones) and steam generator tubes material with and without defects. Characteristics of the defects have been deduced from experience feed back analysis, - Thermo mechanical analysis of the hot leg behaviour and steam generator using both detailed 3D (CAST3M code) calculations and also simplified modelling of hot leg behaviour. Material behaviour modelling is based on experiments results, Statistical analysis of the presence of defects in the steam generator tube taking into account inspection programme results and maintenance programme For RBMK induced damage of fuel channel is part of accident analysis Creep rupture failure of the RCS piping (hot leg and surge line) and SGTR are addressed in the APET. The quantification was based on a set of Sizewell B specific design curves developed for these components (using Larson-Miller approach) and a decomposition event tree based on failure times of these components. Induced SGTR included in APET for PWR. Interfacing LOCA with failure of isolating the break and direct release to the environment is included for BWR. RCS pressure boundary failure is mentioned as important phenomenon Induced failure of hot leg, surge line and SGTR and stuck open power operated reliev valves (PORV) are considered in the APET. Thermo-hydraulics analysed by MELCOR, structural mechanics by expert judgement. Primary system failure assumed according to MAAP (Larson-Miller) calculation to an equivalent of a 100 mm tube. Issue is insignificant because of very low probability for high pressure scenarios. The table shows that induced breaks are taken into account by all organisations involved. However the scope is partly different: While all participants take into account structural failures of the RCS (primary loop and steam generator tubes), two partners mention stuck open safety valves explicitly as a potential release path in case of SGTR. But also other partners might have assumed this path. It is not clear from the comparison whether this is caused by particular valve design. It would be interesting to compare the analysis methods of the partners more in detail,but the available information is too scarce to fully evaluate methods or results. 3. VESSEL FAILURE MODE AND CONSEQUENCES If accident management fails to depressurize the reactor coolant system, and if no induced failure (see previous section) occurs, the reactor vessel failure mode and the related consequences are of particular importance. This is due to the combined effects of thrust forces on the vessel and the melt dispersal. The relevance of these phenomena depends strongly on plant design, in particular on the cavity design. In a narrow cavity with narrow flow paths to

4 4/11 the rest of the containment the rocketing of the RPV may be of particular concern, while a wide cavity which is well connected to the rest of the containment may be more vulnerable to the direct containment heating (DCH) issue. The following table is taken from the SARNET WP5.1 report [1]. AVN EdF Organization FRAM- ATOME GRS IRSN LEI NNC SWED- POWER TUS UJV VEIKI Treatment of the phenomenon in the APET Avoidance of RPV failure by flooding of the cavity DCH, rocket mode failure, pressure rise at vessel failure Combination of calculation and expert judgment for DCH Failure mode assumed according to MAAP creep model; consideration of DCH (simplified approach based on RUPUICUV formula and DISCO experiments for the EPR) and rocket mode failure in case of high pressure failure. As a consequence large (1 m 2 ) break of the containment is assumed. In case of high pressure RCS failure containment failure due to rocketing effect is assumed. For DCH issue a parametric model is linked to the APET. DCH is found to be not significant for Konvoi plants DCH associated with and without hydrogen burn; DCH calculated with ASTEC (RUPUICUV model) taking into account correlations fro KAERI experiments for melt entrainment; geometrical and thermo hydraulic conditions taken into account... Qualitatively considered (not quantified) A detailed assessment of the vessel rocket mode failure was made and shown to be an insignificant event to be excluded in the APET. HPME and DCH were treated as separate nodes. The quantification was supported by information from large scale experiments performed in the US, Sizewell B geometry specific dispersal experiments and calculations from CONTAIN- CORDE and MAAP. DCH, reactor vessel failure, vessel thrust forces assessed based on expert judgment Reactor pressure vessel failure, debris ejection from the vessel and DCH (method?) Vessel rocket based on simplified model and DCH based on expert judgment Failure mode assumed according to MAAP creep model; in case of DCH cavity and containment load was calculated by CONTAIN code; DCH with consequence of temperature induced cavity failure. The table shows that two partners have used the CONTAIN code for DCH based on MAAP initial conditions, one partner uses the RUPUICUV model of ASTEC, while others rely on expert judgment combined partly with simplified calculations. Plant specific experimental

5 5/11 results are also used by two partners: Sizewell B experiments (NNC) and EPR specific DISCO experiments (FRAMATOME). About half of the partners treat the effects of thrust forces either by simplified calculations or by the assumption of containment failure in case of high pressure breach of the RCS. 4. HYDROGEN COMBUSTION Containment failure due to hydrogen combustion depends on the hydrogen risk mitigation strategy (inertisation, igniters, recombiners), the containment size (in relation to the amount of hydrogen) and the containment design (ultimate strength, potentiality of flame acceleration). It seems to be more relevant for the VVER plants than for large dry PWRs. The following table is taken from the SARNET WP5.1 report [1]. AVN EDF Organization FRAM- ATOME GRS IRSN Treatment of Hydrogen issue in PSA level 2 Hydrogen in-vessel generation, Hydrogen concentration in containment, Hydrogen deflagration/detonation and containment failure due to hydrogen combustion are taken into account before, at, and after vessel breach. No further information is given. Hydrogen deflagration/detonation and containment failure due to hydrogen combustion are taken into account before and at vessel breach. No further information is given. Before RPV failure: The likelihood of ignition is taken into account. Basis is the amount of hydrogen because recombiners are seen as the major ignition source. The likelihood of containment failure in case of ignition is evaluated based on deterministic calculations with MAAP4. Monte Carlo calculations are performed for AICC pressure and for the likelihood of fast deflagration varying hydrogen mass, steam concentration and other hydrogen relevant parameters. At RPV failure: Simplistic DCH model takes into account hydrogen generation and combustion. After RPV failure: General considerations indicate that late containment failure due to hydrogen combustion is possible only if no previous burn occurred. No further information is given. Konvoi-PWR: Hydrogen issue has been considered in a simple physical model for all accident phases (including containment atmosphere mixing, passive autocatalytic recombiners, ignition probabilities, AICC model with correction factors) which is linked to the APET. Hydrogen burn is likely in the early phase, but no significant containment challenge. Late hydrogen burns are unlikely due to recombiner action which finally leads to a consumption of all the oxygen. 1300MWe-BWR: This BWR is unique in Germany because it is inerted only in the wetwell. Therefore early hydrogen burns are a significant containment failure reason. The analysis method is comparable to that for the PWR above. (After finalisation of the PSA this BWR has been equipped with recombiners.). Hydrogen burning has been considered at the following times if possible according to physical calculations using ASTEC code (mixture flammable): - core complete uncovery, - beginning of a corium pool formation inside the core, - first corium relocation into the lower head,

6 6/11 LEI NNC SWED- POWER - time of vessel rupture, - time of first fast phase of corium concrete interaction. The probabilities of combustion have been determined by consensus between experts. Experts have considered that: - probability for the first 3 times depends on the availability of electrical sources, - for these same 3 times, in case of presence of electrical sources, the probability of combustion depends on time according to an exponential law which has been adjusted to time of combustion during TMI2 accident, - for combustion at vessel rupture, probability depends on water presence in the cavity pit. The possibility of fast deflagration regime and of Deflagration Detonation Transition (DDT) has been investigated systematically using ASTEC code to assess distribution of gases (multicompartment code CPA) and using a specific simplified fast running combustion code named PROCO for combustion study... Conclusions lead to not consider flame accelerations. Thus the peak pressure has been assessed by ASTEC code simplified model of combustion at the different selected times. For RBMKs, Hydrogen accumulation, damage of the building and support structures in case of hydrogen explosion or large LOCA are considered as key processes for accident analysis. No further information is given. Hydrogen burn is addressed in a decomposition event tree (DET) with the following issues evaluated for four consecutive time frames - the extent of mixing - combustible hydrogen concentration to support hydrogen deflagration or detonation - the presence of an ignition source - the potential of direct transition from deflagration to detonation. When quantifying hydrogen concentrations of the containment, MAAP analyses were used as a source of information, both for hydrogen distribution behaviour in the containment and for hydrogen generation. The mixing data in the DET model are based on expert judgement. Ignition probability was quantified on the basis of availability of AC power. If electrical power is available, the ignition probability is With no electrical power, but having an identified ignition source, the ignition probability is 0.9. With no identifiable ignition source, the ignition probability is 0.5. The quantification of flame induced DDT was based on the Sherman-Berman methodology, taking into consideration the geometry (scale, obstacle, degree of confinement etc.) and mixture conditions. A very low probability of containment threatening combustion events could be demonstrated for Sizewell B with the DET approach summarised above. This was attributed to two aspects: good mixing in the open large dry containment design, and the inerting influence of a steam rich environment... Hydrogen burn and its consequences are covered by nodal questions in the CET for each phase of the accident. No further information is given. Generally, hydrogen combustion is the risk dominating phenomenon for large early releases to the environment both in BWR and PWR. Steam explosion is

7 7/11 TUS UJV VEIKI also considered but gives lower contribution than hydrogen combustion. Hydrogen combustion is mentioned as important phenomenon. No more information is given. Simple physical models and physical parameters for the hydrogen issue are included in the APET for each phase of the accident. The first part of the model is the estimate of in-vessel hydrogen production. The production was sorted to three classes, based on MELCOR results. The next step is the calculation of mole concentration in the accessible containment volume. Four different modes of hydrogen behaviour are assumed in the third step: No burning / Diffusion burn / Deflagration / Detonation. Decision about these modes depends on the mole concentration ranges, existence or recovery of electricity and inertisation against detonation. Deflagration is the preferred mode of combustion even with detonable hydrogen concentration. The final part of the model is the calculation of consequences. When detonation occurs, there is always large containment failure. When deflagration occurs, the pressure peak is calculated from the AICC model and then reduced by reduction factors. The hydrogen model was subjected to large modification between revision 1 and 2 of the PSA, because the initial results indicated only very small risk connected with hydrogen, it contributed to about 2 4 % of core damage frequency (CDF) to early containment failure. The change that contributed most to the increase of the risk of early containment failure to almost 12 % of CDF, was the requantification of several basic events, in particular the probability of diffusion burn versus deflagration in the presence of natural igniters due to electric power. More detailed information on the model is given in the WP5.1 report. Containment pressure without burn, temperature, volumetric gas concentrations were calculated by the MAAP code, then the containment load was determined by using the AICC code, assuming hydrogen ignition at many different points in time. The Sherman-Berman model was used for the determination of DDT conditional probabilities. The probability distribution for the containment load was derived from the loads above and the ignition probability distribution. The probability of the spontaneous ignition in the containment has been taken to be 0,5. This means a cumulative probability for the whole duration of flammable state of the mixture. The igniting of the hydrogen can also by the recombiners (recombiners are provided for DBA hydrogen, not for severe accident mitigation). If the hydrogen concentration exceeds 10 vol.%, then the ignition by recombiners was taken into account, reaching 100% ignition probability if 20% hydrogen exists for 1800s. The treatment of hydrogen combustion differs strongly among the partners. In most cases the analysis is related (in some times limited) to an integral code, such as MAAP4, MELCOR or ASTEC to at least provide the boundary conditions, in terms of amount and release rate of hydrogen and the atmospheric condition (steam concentration). Some partners (5 out of 11) take into account a non-1 ignition probability considering the availability of electrical power, the ignition by recombiners or TMI experience. Most partners consider AICC pressure as a first step for risk assessment (some even as the only one) knowing that deflagration pressure is well over predicted by AICC (some partners

8 8/11 apply a reduction factor to the AICC pressure.) AICC pressure can be either calculated directly with the above mentioned integral codes or by dedicated codes. At least for large dry containments the risk of containment failure due to hydrogen combustion mainly results from the phenomenon of flame acceleration, which may result in DDT. Five out of eleven partners consider this effect in the elevation of the hydrogen risk. This assessment can be done with different approaches: Two partners use the Sherman- Berman model, where rooms of the containment are classified according to their potentiality to support flame acceleration and where the gas mixture is classified according to the reactivity. One partner uses criteria which have been recently developed: Flame acceleration is assessed based on the mixture quality (hydrogen concentration, steam concentration and gas temperature) and a limit value derived from different experiments (sigma criterion). DDT is assessed based on the mixture quality (defined as detonation cell width) and a characteristic length of the room (lambda criterion). Most partners assess hydrogen risk in different accident phases, e.g. before, during and after vessel failure, where conditions for combustion (deflagration, diffusion flame) and probability of ignition may differ. No partner seems to explicitly consider thermal loads from the combustion as a risk for containment failure. 5. DEBRIS COOLABILITY IN EX-VESSEL PHASE After RPV bottom failure the core material enters the reactor cavity. If the debris cannot be cooled permanently there, they will eventually penetrate the containment. Within this process erosion of concrete is of concern because much incondensable gas (part of which is ignitable) is generated, contributing to containment pressurization. It seems to be straightforward to cool the debris by water injection into the cavity. However the success probability of this measure is limited and it has drawbacks in terms of an increased containment pressurisation and of a potential melt-water interaction. It is one of the PSA issues to evaluate this kind of mitigation measure. The following table is an excerpt of the SARNET WP5.1 report. AVN Treatment of ex-vessel debris coolability in PSA level 2 Doel 1/2 APET: Thickness of debris bed in cavity, availability of water injection after vessel breach and basemat melt through is taken into account. Tihange 1: Debris quench in the cavity, fraction of debris participating in core concrete interaction (CCI) and long term debris coolability are sub-trees of the APET. EDF Penetration of foundation raft is determined for three time slots (24h, 48h, 72h) Organization FRAM- ATOME GRS For EPR, late containment failure due to penetration of foundation after failure of melt retention concept is taken into account. Konvoi-PWR: Melt attack at bottom of containment has been considered in detail. Ventilation ducts and details of the sump design to be taken into account. Sump suction lines defined as potential failure location. Later on, gross penetration of foundation is very likely.

9 9/11 IRSN LEI NNC SWED- POWER TUS UJV VEIKI 1300MWe-BWR: Details of the containment design (steel plates at the bottom of the concrete containment) have been identified as potential failure location. Compared to this, other late containment failure modes (pressurisation, penetration of foundation) are insignificant. The APET contains a physical model for CCI. Time of foundation penetration and containment pressurisation are determined. Practically, the results of the physical calculations lead to estimate that the benefit of ex vessel debris cooling was very low when considering the time to basemat meltthrough and has not been considered. On the contrary, ex vessel debris cooling by water lead to an increase of the containment pressurisation which is taken into account (earlier time for the filtering/venting U5 device opening, and eventually containment damage due to over pressurisation). For RBMKs, reactor core consists of fuel assemblies, which are located in the fuel channels. Limited information is available about debris behaviour in ex-channel phase. CCI and basemat penetration are taken into account in the CET. CCI and basemat penetration are taken into account in the CET. CCI and related gas production are taken into account. Rupture or leak of the cavity door due to core material impact or radiation heating is an important issue. CCI and basemat failure are taken into account in the CET. These phenomena are influenced by water presence in the cavity and at the same time lead to loss of this water. Cavity flooding with ECC water coming through the failed vessel is also taken into account. With the present plant configuration, the effect of this flooding is only temporary, late recirculation cannot be reached as this water escapes through ventilation lines and bypasses the recirculation sump. In low pressure cases, when pressure does not represent a challenge to cavity integrity, the temperature of the cavity door may rise to 450 C. At this temperature the door seal may be damaged and the containment leakage increases. In the evaluation of APET sequences with low-pressure RPV failure the increased leakage (with 0.01 m2 leakage surface) is taken into account. In high-pressure sequences the door would be damaged because of high temperature, but for these cases cavity damage is anyhow taken into account because of the high pressure. Further MCCI including gas generation is analysed by MAAP. The table shows that ex-vessel debris behaviour is taken into account by all organisations involved. The information given by the organisations has a low degree of detail. From the available information one can conclude that MCCI, penetration of the foundation and associated gas generation are the most important concern. However no information about the models to evaluate this issue has been given. Some organizations mention as relevant issues melt attack on cavity doors and their seals, on steel walls of the containment, and on sump suction lines. These potential failure modes are

10 10/11 highly design specific with the tendency to earlier fission product release compared to basemat penetration. If they occur, they are much faster than concrete erosion. 6. IMPROVEMENT NEEDS Improvement of the quantification of physical phenomena should focus on those phenomena that govern the risk of the plant. For most types of plants the risk seems to be dominated by early large release of fission products. Three main phenomena contribute, in case of a PWR, to early large release: steam explosion, hydrogen combustion and the phenomena in the context of RCS failure under high pressure. As there seems to exist an international consensus that in-vessel steam explosion with the consequence of containment failure is extremely unlikely and as hydrogen related issues are well understood and well describable with the exception of ignition probability, see below improvement needs are related to the high pressure issues. Assessment of the RCS depressurization and the quantification of the consequences in case of high pressure breach are of particular importance. High pressure issue (induced RCS failure and RPV failure mode and consequences): Assessment of passive RCS failure, both within the containment and as SGTR leading to containment bypass, is a straight forward process. However there is a complex dependency between the kind of core melt progress, the temperature and pressure evolution in the primary components, the four main depressurisation modes (active depressurisation by staff / stuck open valve / failure of hot leg / failure of steam generator tubes), and possible recovery of low pressure RPV injection. It seems to be very difficult to address this issue properly by ordinary event tree models. It requires a large amount of calculations for different pressure - and temperature histories with standard codes, such as MAAP4, that have a state-of the art creep model, in order to create probabilistic numbers.. One approach may be a set of deterministic calculations governed by a suitable probabilistic overhead. This kind of analysis, known as dynamic event trees has been applied by one of the partners and is subject of further SARNET activities. While all institutions take into account induced breaks due to failure of pipework (primary piping, steam generator tubes), only a minority addresses the issue of stuck open safety valves, because of the poor experimental data base. According to the GRS experience, the probability for stuck open valves under core melt conditions may be significant and contributes to depressurisation of high pressure scenarios. The assessment of the consequences of the RPV failure under high pressure is also complex. Two different failure modes of the containment are imaginable: a mechanical failure due to missiles from the RPV thrust forces and thermal (resp. pressure) loads from a fraction of melt that is dispersed in the containment atmosphere (DCH). Both phenomena occur at the same time and interfere: the possible movement of the RPV changes the flow conditions for DCH. Next step, and one of the most difficult one, is then to determine possible damage of the containment in terms of break size in order to quantify the source term. The hydrogen issue is taken into account by all organisations involved, and in many cases it is an issue with relevance for the overall PSA result. From the available information one can conclude that there is no common approach to determine the ignition probability, which is a crucial issue, and the existing approaches seem to be very different and not well founded.

11 11/11 Since this is a topic which is almost independent of the plant design, there seems to be a good chance to address this in an international cooperation. The calculations on the atmospheric conditions (hydrogen distribution, steam concentration etc) mostly rely on lumped parameter codes. Their results depend to a significant degree on the nodalisation of the volumes involved. It seems that partners use very different degrees of nodalisation. It would be useful to establish recommendations on the necessary degree of nodalisation, possibly supported by other methods, such as CFD analyses. Ex-vessel debris behavior and coolability of the melt is difficult to assess because analytical models are in a mediocre state and the direct assessment of experimental data is difficult and controversy. Molten core concrete interaction, and associated gas generation are the most important concern. But, on the other hand, for most plants basemat penetration is not riskrelevant. Improvement of the situation might be achieved when the ongoing research on CCI and ex-vessel coolability can be incorporated into the PSA analyses. It seems that plant specific details (concrete composition, existence of drains and ducts in the concrete, doors, sealings, penetrations) may have significant impact on the ex-vessel debris behaviour and on containment vulnerability. Conclusions from one plant to another should not be drawn unless the comparability is proven. A general issue that needs continued improvement is the incorporation of integral code results in the process of APET quantification. On the one hand side, faster running (and perhaps more simplified) codes are needed to analyze a large number of sequences and their potential variations, and on the other side the inclusion of CFD codes for the assessment of special purposes must be envisaged. Some partners use Monte Carlo technique to generate branch probabilities following code calculations either with simplified stand-alone codes or with parts of the integral codes. One reason for asking for fast running integral codes is the application of the Monte Carlo technique for the assessment of the entire scenarios in the context of dynamic event tree analysis mentioned above. 7. REFERENCES [1] Comparison of partners methodologies for level 2 PSA development IRSN/Reactor Safety Division FT/DSR/SAGR/ (June 2005)