State of the Art and Challenges in Level-2 Probabilistic Safety Assessment for New and Channel Type Reactors in India Abstract

Transcription

1 State of the Art and Challenges in Level-2 Probabilistic Safety Assessment for New and Channel Type Reactors in India R.S. Rao, Avinash J Gaikwad, S. P. Lakshmanan Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai, India Abstract Level-2 Probabilistic Safety Assessment (PSA) deals with the accident scenarios involving significant core degradation or core damage due to severe accidents, which are identified in Level-1 PSA. This provides insights into the relative importance of accident sequences leading to core damage in terms of the severity of the releases of radioactive material they might cause. It also provides insights into weaknesses in measures for the mitigation and management of severe accidents and ways of improving them. An important part of Level-2 PSA is modelling the phenomena that occur between the onset of core damage and containment failure. Developing understanding of these phenomena is necessary to quantify the probability of failure of the containment and the source term. The uncertainties involved in the severe accident progression may impact the results of the Level-2 PSA substantially. Detecting the onset of severe accident will also help in severe accident management guidelines (SAMGs) development with proper provisions and special instrumentation. Severe accident phenomena are highly complex and in spite of the extensive studies there are still many uncertainties involved in the accident progression prediction as limited amount of experimental data is available. Also the experiments are carried out with simplified separate effect test and involve lack of repeatability. Severe accident progression in horizontal channel type reactors is different from light water reactors till the corium debris bed is formed at the bottom of the calandria vessel. Presence of moderator in the channel type reactor slows down the core disassembly significantly. There are also significant differences in corium composition and corium geometry. As mentioned above, the severe accident progression forms the major input to the Level-2 PSA quantification and accuracy of results. This paper aims to bring out the uncertainties involved in the accident progression for pressurized heavy water reactors. Some of the important issues such as deformation of pressure and calandria tubes, core disassembly, steam supply into the interior of debris, hydrogen generation during relocation of melt, hydrogen deflagration and detonation, possibility of steam explosion, failure of channels under certain conditions like stagnation channel break and non-uniform temperature distribution and its impact on other channels need to be resolved for use in Level-2 PSA. The issues once resolved would significantly improve the PSA models and produce realistic results (Source term). Design changes and incorporation of provisions for severe accident management especially after Fukushima would also be required to be included in the Level-2 PSA models. In India, currently two VVERs are under construction and in advanced stage of completion. First of a kind systems like core catcher, beyond design basis hydro-accumulators and passive heat removal system are incorporated in these reactors. Passive autocatalytic recombiners and 1

2 primary depressurisation systems are also installed. These systems would reduce the likelihood of large early releases by avoiding certain phenomena such as direct containment heating, hydrogen detonation etc. Consideration of these systems in Level-2 PSA substantially reduces the consequences. The paper highlights all these aspects of Level-2 PSA. 1.0 Introduction Severe accident progression in horizontal channel type reactors is different from light water reactors till the corium debris bed is formed at the bottom of the calandria vessel. Presence of moderator in the channel type reactor slows down the core disassembly significantly. Figure 1 shows the schematic of a typical light water reactor and Figure 2 shows the schematic of a PHWR. There are also significant differences in corium composition and corium geometry [1, 2, 3]. Severe accident phenomena are highly complex and in spite of the extensive studies there are still many uncertainties involved in the accident progression prediction as limited amount of experimental data is available [4, 5, 6]. The severe accident progression and assumptions made form the major input to the Level-2 PSA quantification and accuracy of results. This paper aims to bring out the uncertainties involved in the accident progression for pressurized heavy water reactors. Some of the important issues such as deformation of pressure and calandria tubes, core disassembly, steam supply into the interior of debris, hydrogen generation during relocation of melt, hydrogen deflagration and detonation, possibility of steam explosion, failure of channels under certain conditions like stagnation channel break and non-uniform temperature distribution and its impact on other channels need to be resolved for use in Level- 2 PSA. The issues once resolved would significantly improve the PSA models and produce realistic results. Design changes and incorporation of provisions for severe accident management especially after Fukushima would also be required to be included in the Level-2 PSA models. In India, currently two VVERs are under construction and in advanced stage of completion. First of a kind systems like core catcher, beyond design basis hydro-accumulators and passive heat removal system are incorporated in these reactors. Passive autocatalytic recombiners and primary depressurisation systems are also installed. These systems would reduce the likelihood of large early releases by avoiding certain phenomena such as direct containment heating, hydrogen detonation etc. Consideration of these systems in Level-2 PSA substantially reduces the consequences. The paper highlights these aspects. 2

3 Figure 1: Schematic of LWR Figure 2: Schematic of PHWR 2.0 Challenges Associated with Level-2 PSA for PHWRs The Pressurised Heavy Water Reactor (PHWRs) severe accident progression is different from the light water reactors till the corium debris formed in the calandria vessel [7-10]. The issues till the corium debris formation in the calandria vessel are specific to PHWRs and more research is required for resolving the issues. The phenomena in the later phase are similar to the light water reactors (LWRs) and the research outcome for LWRs is generally applicable for PHWRs also in this phase. The issues on which more work is required are brought out in this section. Issue 1: Core disassembly: The core damage progression of PHWR is different from that of light water reactor in a severe accident. The following are the issues to be resolved Pressure tube ballooning and sagging. PT and CT contact heat transfer coefficient. Development of core disassembly model Experimental/theoretical studies are required. Material properties are required at high temperatures. Corium composition and corium geometry. Thermal-mechanical behaviour: deformation of the PT and CT, hydrogen production etc. Experiments have being performed for low modest temperature ranges. However, in severe accidents, much higher temperatures would be seen. Therefore experiments should be performed for understanding the thermal, mechanical and chemical behaviour. 3

4 Experiments were carried out in AECL Chalk River laboratories to investigate the core disassembly process. However, the oxidizing environment on channel disassembly is not studied. This should be studied. Figure 3: Core disassembly Pipe whip Figure 4: Phenomena of pipe-whip inside the vessel 4 Jet & hot material impingement Expanding steam bubble Shattered and dispersed fuel

5 Issue 2: Uncertainties in Heat and Hydrogen Generation During the accident progression, debris are collected at the bottom of the calandria vessel. Steam supply into the interior of debris is a source of uncertainty. This will result in uncertainties in heat and hydrogen generation in metal water reaction. Issue 3: Material - Moderator Interaction A sudden drop of a large mass of hot, molten/solid material to the bottom portion of the vessel would lead to arising of strong steam surges as the stored heat gets transferred to water instantly. Pressurisation of calandria vessel may take place because of sudden steam surges. Even with the large relief flow area, this pressurization could challenge the integrity of the calandria vessel. Experimental studies are required. Issue 4: Ex-vessel debris cooling Steam explosion: When calandria vessel fails, the debris relocates into the calandria vault where relatively cold water may be present. This may lead to steam explosion. There are uncertainties involved in this phenomena with respect to possibility of occurrence of steam explosion and its efficiency. Coolability: Some experiments showed that the corium may not be coolable beyond certain depth from the overlaying pool of water. This also depends on the porosity and composition of the debris. Issue 5: Re-flooding of a damaged core When the core is hot, the flooding of core with water may actually lead to further metal water reaction and heat-up instead of cooling. This issue gained importance after TMI-2 and reconfirmed in the recent Fukushima accident. Some experiments were carried out internationally in this regard (e.g. CORA experiments). The possibility of such phenomena and its deterrent effects in PHWRs may be studied. Issue 6: Calandria vessel failure at high heat fluxes It is generally believed that when the corium slumps to the lower portion of the calandria vessel, heat gets transferred to the surrounding calandria vault water by conduction and convection. However, under high heat flux conditions, there is a possibility of forming a vapour blanket on the outer side of the calandria vessel resulting into the heat up and failure much before significant amount of calandria vault water boils off. 5

6 Issue 7: Fuel channel deformation (Sagging/ballooning) The fuel channel is treated as failed when both PT and CT fail. Non-uniform circumferential temperature distributions lead to pressure tube ruptures. Subsequently, CT also will fail due to impingement of hot steam from the primary heat transport system, if the accident scenario progresses at high pressure (e.g. SBO). (Also refer to Issue 12) During the accident progression, the PT and CT may fail at high pressures. Interaction of material with the neighboring channels needs to be studied. Fuel channels also may fail because of a local melt-through or sagging of pressure and calandria tubes under low pressures. Experimental studies are required for developing the models. Issue 8: Hydrogen mitigation Large amount of hydrogen is released during the accident conditions and pose a threat to the containment integrity. At present, the mitigation does not even exist for the LOCA + LOECC (design basis accident for PWHRs). In case of severe accidents much large hydrogen is released and definitely lead to early failure of the containment. Development and deployment of passive recombiners for mitigation of severe accidents are beneficial. Understanding of the local and global behavior of hydrogen distribution. Issue 9: Molten Corium Concrete Interaction Molten material react with the concrete basemat and this produces steam, non-condensable gases such as carbon dioxide and combustible gases such as hydrogen and carbon monoxide. Models need to be developed and validated for understanding of these phenomena. Concrete properties also need to be established. Issue 10: Thermal ablation of the opening of the calandria vessel When calandria vessel fails, molten material ejects out from the vessel to the vault. Amount of debris ejection depends on the size of the opening of the calandria vessel. The size of the opening may increase by the ablation of molten material. This needs experimental studies. 6

7 Issue 11: Debris Layer Formation Experimental studies are required for better understanding of Separation/stratification of core debris layers i.e., oxide layers and metal layers. Studies are also required for formation of crusts and their rupture. Issue 12: Stratification Stratification of steam and water in the coolant channels (single/multiple channel (s)) may occur during the accident scenarios. This may lead to failure of the PT and CT even before channel uncovering from the outside (moderator side). Assumptions/simplifications are made in the current Level-2 PSA performed. Detailed investigations are required to resolve these uncertainties to arrive at the realistic Level-2 PSA results. Recently the core damage definition for PHWRs is standardized for the PHWRs. The definition of large early release frequency also need to be standardized. 3.0 Positive Aspects of VVERs The current generation of VVERs are being built with advanced features such as core catcher, passive autocatalytic recombiners (PARs), passive heat removal system, second stage hydroaccumulators, quick boron injection system, primary depressurization system etc. The PARs combine the hydrogen with the oxygen and in turn the hydrogen concentration is reduced in the containment and decrease the probability of containment failure because of hydrogen detonation. This eliminates the early failure of the containment during the accident progression due to hydrogen detonation. Similarly, the primary depressurization system operates when the core exit temperature exceeds a certain value which depressurize the primary system and the failure of reactor pressure vessel during the accident progression would be at lower primary pressure, thus this feature practically eliminate the high pressure melt ejection and direct containment which would in-turn eliminates the large early during the accident progression. Core catcher would collect the molten material and cool it. This would eliminate the molten corium concrete interaction (MCCI), hydrogen generation during the MCCI, other combustible and non-condensable gases generation. This would further reduce the failure probability of containment. Passive heat removal system and second stage hydro-accumulators together would be able to remove the decay heat during the loss of coolant accident with station blackout conditions for reasonable amount of time. After which, recovery actions are possible to remove the decay heat. 7

8 The quick boron injection system injects boron into the primary passively with the help of differential pressure across the pump. This system is designed for anticipated transients without scram. This would further reduce the probability of core damage and containment failure in the events in which reactor power excursions takes place. The above features practically eliminate the large early releases and also reduce the likelihood of containment failure in the long term during the various accident scenarios. These features would substantially reduce the large early release frequency for VVERs. 4.0 Conclusions Severe accident phenomena are highly complex and involve large uncertainties in the accident progression prediction. Severe accident progression in horizontal channel type reactors is different from light water reactors till the corium debris bed is formed at the bottom of the calandria vessel. Presence of moderator in the channel type reactor slows down the core disassembly significantly. There are also significant differences in corium composition and corium geometry. As mentioned above, the severe accident progression forms the major input to the Level-2 PSA quantification and accuracy of results. Some of the important issues such as deformation of pressure and calandria tubes, core disassembly, steam supply into the interior of debris, hydrogen generation during relocation of melt, hydrogen deflagration and detonation, possibility of steam explosion, failure of channels under certain conditions like stagnation channel break and non-uniform temperature distribution and its impact on other channels need to be resolved for use in Level-2 PSA. The issues once resolved would significantly improve the PSA models and produce realistic results (Source term). Design changes and incorporation of provisions for severe accident management especially after Fukushima would also be required to be included in the Level-2 PSA models. In India, currently two VVERs are under construction and in advanced stage of completion. First of a kind systems like core catcher, beyond design basis hydro-accumulators and passive heat removal system are incorporated in these reactors. Passive autocatalytic recombiners and primary depressurisation systems are also installed. These systems would reduce the likelihood of large early releases by avoiding certain phenomena such as direct containment heating, hydrogen detonation, steam explosion etc. 5.0 References 1. INTERNATIONAL ATOMIC ENERGY AGENCY, Analysis of Severe Accidents in Pressurised Heavy Water Reactors, IAEA TECDOC-1594, IAEA, Vienna (2008) 2. R. S. Rao, Avinash J Giakwad, Presentation in Severe Accidents Theme Meeting, SRI, Kalpakkam, India (2012) 3. S. K. Gupta, Presentation in Severe Accident Theme Meeting, AERB, Mumbai, India (2007) 8

9 4. B.R. Sehgal, Accomplishements and Challenges of the Severe Accident Research, Nuclear Engineering and Design 210, pp (2001) 5. Jean-Pierre Van Dorsselaere, T. Albiol, Jean-Claude Micaelli, Research on Severe Accidents in Nuclear Power Plants, Nuclear Power Operation, Safety and Environment, IRSN, France. 6. D. Magallon et. al., European Expert Network for the Reduction of Uncertainties in Severe Accident Safety Issues (EURSAFE), Nuclear Engineering and Design 235, pp (2005) 7. D. Duplec, M. Mladin, I. Prisecaru, Generic CANDU6 Plant Severe Accident Analysis Employing SCADAPSIM/RELAP5 Code, Nuclear Engineering and Design, 239 (2009) 8. G. Negut, A. Catana, I Prisecaru, D. Duplec, Severe Accident Development Modelling and Evaluation for CANDU, Annals of Nuclear Energy 36, pp (2009) 9. M. Mladin, D. Duplec, I. Prisecaru, Modification in SCDAP Code for Early Phase Degradation in a CANDU Fuel Channel, Annals of Nuclear Energy 36 pp (2009). 10. Thinh Nguyen, Development of Severe Accident Management Guidance (SAMG) for the Canadian CANDU 6 Nuclear Power Plants, Nuclear Engineering and Design 238 pp (2008) 9