ASTEC application to HTRs

Transcription

1 1) PBMR, Pretoria (RSA) ASTEC application to HTRs A. RAMLAKAN 1), N.NAIDU 1), M. SANYASI 1), D. NAIDOO 1) ABSTRACT The initial IRSN-GRS requirements for the development of the ASTEC European integral code covered all present and future PWR, VVER and BWRs. At present the ASTEC code is fully applicable to all severe accident scenarios, at power operation, of Generation II PWR and VVER. This paper discusses present and potential ASTEC application to HTRs with specific focus on the use of ASTEC at the Pebble Bed Modular Reactor company. A description of present HTR designs is given and a gap analysis is presented in terms of the present ASTEC capabilities and HTR needs. The Verification and Validation status of the ASTEC code is also discussed. The simulation of the transport of graphite dust and radionuclides in the reactor building following a 4mm and 100mm pipe break, located in the Reactor Top Cavity, was performed using the ASTEC CPA containment module. The evolution of these accident scenarios is described in this paper. These analyses demonstrate how ASTEC can already be used in HTR analysis and also serve as examples to explain the areas of the code that need improvement and validation. 1. INTRODUCTION The South African Pebble Bed Modular Reactor (PBMR) is a High Temperature Gas-Cooled Reactor (HTR) with a pebble bed core. The proposal presented for the PBMR design is a vented and filtered confinement system rather than a conventional containment system. The differences between a conventional LWR containment and an HTR confinement are not only limited to different engineering structures but also to differences in source terms, i.e. gas, aerosol, fission product composition and gas pressures and temperatures. One of the interim tasks performed by PBMR as part of the United States Department of Energy Next Generation Nuclear Plant (NGNP) pre-conceptual design were plant level assessments in support of radionuclide retention allocations for the fuel, helium pressure boundary and reactor building. Two scenarios of 4 mm and 100 mm pipe breaks in the Reactor Top Cavity must be analysed with suitable models of the ASTEC European integral code [2] as part of a program of analyses that requires data from various codes. The ASTEC CPA containment module has been chosen as it is a lumped parameter code, originally developed in the field of Light Water Reactor (LWR) containments, but which includes model and features which enables it to properly simulate HTR confinement scenarios. Previous modelling of HTRs using ASTEC has been detailed in [3]. Alternative codes which can be considered for HTR confinement simulation are COCOSYS from GRS (Germany) [4], MELCOR [5] and CONTAIN [6] from USNRC (USA). If an accident in the form of a pipe rupture should occur, the Nuclear Heat Supply System (NHSS) will take time to depressurise. The FLOWNEX [7] code, developed by M-Tech Industrial and used at PBMR, provides the thermal-hydraulic input [8] into the ASTEC model which includes the mass flow rate, temperature and pressure of the incoming gas as a function of time. Radionuclide and dust source terms are provided as a mass flow rate or cumulative mass injected as a function of time. For larger breaks, typically 100 mm to 1000 mm Single Ended Guillotine Breaks (SEGB), the depressurisation will be very quick. For smaller breaks, typically from 1 mm to 10 mm SEGB, there will still be some flow in the primary coolant system of the NHSS when decay in the Session 2 : "ASTEC", paper 2.1 1/19

2 reactor starts and this acts as a mechanism to carry the delayed releases into the building. The depressurisation times for the various break sizes are of importance in determining the radionuclide release from the NHSS reactor building. However, even after a total initial depressurization, the helium inventory still left in the system will slowly heat up due to decay heat from the reactor. This will cause expansion of the gas which pushes out some of the inventory through the break into the Pressure Relief System (PRS) of the Reactor Building (RB). Radionuclide and dust masses as well as thermal-hydraulic results from the analysis for all the compartments of the ASTEC reactor building model are presented in this document. The radionuclides modelled are Ag-110m, Ag-111, Cs-137, I-131, I-133, Sr-90 and Te-132. These radionuclides are considered the dominant contributors to the offsite TEDE and CEDE [9] based on insights from previous HTR studies [10]: - The CEDE is defined as the Committed Effective Dose Equivalent which is the sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues. - The TEDE is the Total Effective Dose Equivalent (TEDE) which is the sum of the effective dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures). These radionuclides have components both attached to dust and not attached to dust. These components are modelled differently and thus the results presented in this document are specified for each component. Radionuclides and dust masses are subdivided per compartment and into the suspended and deposited mass. The thermal-hydraulic data are the temperatures and pressures in each compartment as well as mass flow rates between compartments. 2. ASTEC SUITABILITY TO HTR 2.1 HTR Description The High Temperature Reactor is a Generation IV reactor concept that uses a graphitemoderated nuclear reactor with a once-through uranium fuel cycle. The reactor core can be either a prismatic block or a pebble-bed core. The HTR system is designed to be a highefficiency system that can supply process heat to a broad spectrum of high-temperature and energy-intensive, non-electric processes. The system may incorporate electricity generating equipment to meet cogeneration needs. The system also has the flexibility to adopt uranium/plutonium fuel cycles and offer enhanced waste minimization. Thus, the HTR offers a broad range of process heat applications and an option for high-efficiency electricity production, while retaining the desirable safety characteristics offered by modular high-temperature gascooled reactors. Significant differences exist between modelling of the source term between a conventional LWR and an HTR. The inherent safety features of modular HTRs make events leading to severe core damage highly unlikely and constitute the main differentiating aspects compared to LWRs. Unlike an LWR, the primary safety barrier for an HTR is performed by the fuel. In the event of a total loss of coolant the fuel temperatures in an HTR do not exceed the fuel design limits and thus the particle coating remains intact. In an HTR, helium is used as a coolant which has important implications in terms of containment design. The coolant in an HTR is thus non-condensable as compared to steam in a LWR which condenses in the containment, thus reducing the pressure in the containment. There is no need to retain the coolant to cool the core as in LWRs. A release from the system will contain radioactive contaminants in the form of circulating and re-suspended dust, circulating activation and fission products and desorbed radionuclides. Session 2 : "ASTEC", paper 2.1 2/19

3 2.2 Module Applicability Due to significant differences between PWR and HTR cores none of the ASTEC models for the core have, at present, any applicability to HTRs. New models will thus need to be written to enable ASTEC to model HTR cores. Figure 1 : Schema of the ASTEC V1.3 Modules At present the CESAR and SOPHAEROS have not been used to model accidents in HTRs. CESAR has been extended to model any number of non-condensable gases in Version 2.0 [11] and thus, after validation for helium, should be able to model the thermal-hydraulics of the NHSS. The aerosol and radionuclide physics in a HTR are similar to that of a LWR with the exception of chemical reactions associated with water and thus SOPHAEROS should be able to simulate the aerosol and fission product transport in a HTR Reactor Coolant System. The CPA module has been used for modelling of an HTR reactor building for different pipe break scenarios. A table of US NRC NGNP PIRT items taken from [12] is presented in Table 1 showing the phenomena which are important to be modelled and the current status of modelling to be found in codes such as ASTEC. A scoping study on a confinement approach to a PBMR type plant [3] was performed to quantify its performance in terms of retention and thermal-hydraulics in case of a postulated breach in the Helium Pressure Boundary which included a code comparison exercise between the ASTEC and CONTAIN codes. The results obtained indicate that both codes predict very similar thermal-hydraulic responses of the confinement both in magnitude and timing. As for the dust aerosol behaviour, both codes predict that most of inventory coming into the confinement is eventually depleted on the walls and only about 1% of the aerosol dust is released to environment. The cross-comparison of codes states that largest differences are in the inter-compartmental flows and the in-compartment gas composition. The ASTEC CPA models for the mass flow include two mechanisms: convection, driven by pressure gradients, and diffusion, generated by differences in concentration of each gas component. However, in CONTAIN the flow is estimated only by convection. The differences in modelling are visible under low gas flow rates where the codes predict different gas compositions in compartments. Session 2 : "ASTEC", paper 2.1 3/19

4 Table 1 : US NRC NGNP PIRT (Confinement Phenomena) Issue (phenomena, process, geometry condition) Importance for NGNP (High, Medium, Low) Level of knowledge (High, Medium, Low) Status of FP modelling (adequate, minor modifications, major need) Confinement aerosol physics High Medium Minor modifications Radiolysis effects in confinement High Medium Minor modifications Filtration High High Adequate Combustion of dust in confinement High Medium; Low Major need NGNP-unique leakage path beyond confinement High High Minor modifications Confinement leakage path, release rate through penetrations High Medium Adequate Pressure-relief-valve filter Low High Adequate 2.3 HTR Reactor Building Description The containment for nuclear facilities protects the public against uncontrolled exposure to activity and is conventionally a low leakage barrier that can withstand a high differential pressure. For HTRs, a conventional high pressure, low leakage containment may not significantly reduce public exposure in an accident any better than a vented filtered containment. In an HTR the use of a conventional containment would result in a high pressure build-up raising the possibility of failure of the containment and leading to a far more severe release. The concept of a controlled vented, filtered containment, however, does have several advantages in this regard: - The source term to the environment is conducted through an inner pathway designed for that purpose; - No carrier gas is available in the core to transport fission products out of the Helium Pressure Boundary (HPB) when fuel temperatures reach their highest values during the accident; - The probability of failure of the containment/confinement is reduced by the relief of pressure and, no less importantly, - Earlier access to the building may be possible in order to close the HPB leak. Due to the designs of HTRs being very different from those of LWRs, where the possibility of core meltdown has largely driven the safety philosophy, a different approach to modelling the safety case is necessary. The PBMR confinement system is defined as the combination of structures and active or passive systems that serve to prevent or mitigate the uncontrolled release of radioactive material to the environment. The confinement system comprises of the following major elements: - Reactor Building structure, including doors and penetration seals; - Heating Ventilation and Air-conditioning (HVAC) system; - Pressure Relief System (PRS). Session 2 : "ASTEC", paper 2.1 4/19

5 The HVAC filters and conditions the air in various ventilation zones within the reactor building during Normal Operation and Anticipated Operational Occurrence (AOO). The PRS limits damage to the elements of the confinement as a result of an overpressure, caused by gas escaping from the pressure boundary, by passively venting the gas out of the building during a depressurization event. The release through the PRS thus can be filtered. After depressurization the confinement is closed to limit gas exchange between the reactor building and the atmosphere. The Post Event Cleanup System (PECS) is initiated after the temperature of the building has cooled down to specified levels. Figure 2 depicts the flow path through the compartments within the Power Conversion Unit (PCU) portion of the building confinement out to atmosphere. The Block Flow Diagram compartments are named according to the equipment which they house. The rupture junctions can be seen, for example between the RTC compartment and the IHX compartment. Other junctions are open, i.e. modelled as atmospheric junctions. The stack is a vertical flue providing a pathway for gas to be released at a stationary point at some height above the building so as to increase air dispersion [19]. Figure 2: Reactor Building Block Flow Diagram A CPA input deck for the HTR reactor building described in Figure 2 was developed. All compartments were modelled using the NONEQUIL option in ASTEC CPA which allows the zone to be separated into a gas and liquid part, openings between compartments were modelled as atmospheric junctions, rupture panels as rupture junctions and the opening to atmosphere as an atmospheric valve with event structures defined for the opening and closing of the junction (an event is the possibility to modify some variables of the data base as a function of any user conditions). Thirteen zones were modelled: 10 for the building compartments, 1 environment zone for release from the stack, 1 environment zone for release from the upper leakage junctions and another environment zone for release from the lower leakage junctions. Over 50 walls were modelled, all using the PROGRESSIVE nodalization option in CPA: using the progressive arrangement of layers in a wall, where the thickness of each layer is increased by a user provided factor from the outer side to the centre or the adiabatic side of the material. The input deck was over 4000 lines long. It is to be noted that this is a simplified representation of the reactor building and that a full representation of an HTR reactor building needs approximately lines with approximately 130 compartments, 1200 walls and 240 junctions. Session 2 : "ASTEC", paper 2.1 5/19

6 2.4 Verification and Validation status The Verification and Validation (V&V) data existing for ASTEC and applicable to HTRs need to be identified from experiments performed during the EVITA [14], SARNET [15] and SARNET2 [16] European projects and compiled in a separate report. For instance, some efforts on CPA validation have been summarised in the reference [17]. This should underline what V&V data is still needed to address the modelling of important phenomena. For example, it was concluded that certain correlations, e.g. the correlations for phoretic deposition mechanism or heat transfers in the containment, already account for the presence of helium but some have only been validated in an air atmosphere. As already written above, further validation is necessary on the extension of CESAR to modelling helium flows, as well as on suppression pools modelling. 2.5 Proposed improvements to ASTEC for HTR adaptation The first version of the new ASTEC V2 series has been released in July 2009 by IRSN and GRS. Several improvements will be included in these versions, some based on the feedback of the ASTEC V1 applications in previous years, others based on the evolution of knowledge following from current international R&D. Some proposed improvements below will partly be answered by model evolutions in the ASTEC V2 versions Modelling improvements Several needs which entail making improvements in the ASTEC V1 models have been identified as an outcome of the present work. They are detailed below: - Dust resuspension in the circuits. - Chemical reactions kinetics capabilities in both the primary circuit and in the confinement. This may be important in regions where temperatures are too low for chemical equilibrium to be considered. - Helium flows in the primary circuit. The extension of CESAR to any type of gas was performed at IRSN in the ASTEC V2 frame but some validation work will be necessary in the near future. This extension may be useful in order to allow tightly coupled direct calculations between CESAR and SOPHAEROS. - Specific safety systems. HTRs may have safety systems different to those in LWRs. - DRASYS model (in the CPA module) for suppression pool computations Numerical issues Some numerical problems that were encountered in this present study are closely linked to the large uncertainties that remain on the knowledge of dust mass in HTRs in general, in particular on resuspension modelling. These uncertainties often lead to the use of conservatism in the quantity of the dust mass resuspended and the time over which this dust mass is resuspended, which often results in a source term to be modelled that may be larger than realistically expected. This may create potential numerical convergence challenges when dealing with the resultant stiff differential equations which are a consequence of mass flows that are very large for a few seconds and then drop suddenly to zero. In ASTEC V1, most problems were solved by using a suitable management of numerical time-steps but investigations should continue. Another numerical issue concerns the fission product concentration in the primary circuit. Due to the fact that the TRISO fuel plays the role of the primary safety barrier, the fission product concentration in the primary circuit is some orders of magnitude lower than that expected to be found in the primary circuit of a LWR during a Loss of Coolant Accident. The order of magnitude may even be lower than the minimal concentrations managed by ASTEC. Investigations should also continue on this issue. The numerical robustness of the suppression pool models should be improved in the future as the code crashes when running some of the DRASYS models. Session 2 : "ASTEC", paper 2.1 6/19

7 2.5.3 Tools for users support For HTRs it would be helpful for the code to use a Graphical Users Interface (GUI) for implementation of compartment data as HTRs have considerably more rooms than LWRs and inputting all geometrical structures manually is time intensive. HTR confinement input decks may have more than lines which significantly impacts error checking and thus an efficient error checking system is needed. A new pre-processing tool, JADE, including a GUI, is implemented in ASTEC version V2.0. It will help the user prepare the input deck and to make easier the checking of complex input decks. It may be important to define guidelines for the users to model HTRs in terms of nodalization for small breaks to allow simulation of possible helium stratification. 2.6 Conclusions on ASTEC applicability to HTR At present only the out of core modules, i.e. CESAR, SOPHAEROS, CPA and IODE, have applicability to HTRs. Extensive validation and verification work is however necessary for these modules in the domain of HTRs. Improvements to the pre and post processing interfaces are also needed. 3. ASTEC REACTOR BUILDING BREAK ANALYSIS 3.1 4mm Break Study Phase 1 (t < 45 minutes) Rupture Panels Break At time of the break the helium, graphite dust (mean diameter 0.1 µm [18]) and radionuclide release from the Helium Pressure Boundary (HPB) to the Reactor Top Cavity (RTC) compartment begins. The RTC compartment begins to pressurize and the temperature and the helium mass fraction in the RTC compartment increase. The temperature increase is relatively sharp in the RTC compartment compared to the remainder of the compartments. Rupture panels along the vent path burst successively as flow of helium takes place and compartments pressurize. The rupture panel between the Reactor Top Cavity (RTC) compartment and IHX compartment opens at approximately 1 minute (60 s), the rupture panels between the IHX and DVSW compartments open at approximately 14 minutes (840 s) and the rupture panels between the SIP compartment and the environment open at approximately 35 minutes (2100 s) which results in helium, radionuclides and dust transport to the environment. All reactor building compartment pressures drop to values close to atmospheric pressure. All reactor building compartment pressures show differences with the atmospheric pressure of the environment due to the contribution of the hydrostatic head term,, with the density of the gas composition in the compartment and the height of the centre of the compartment. Session 2 : "ASTEC", paper 2.1 7/19

8 Figure 3: Zone Pressure Evolution for the First 45 Minutes of Depressurization Phase 2 (t < 91 hours) Completion of Blowdown Once the final rupture panel to atmosphere has opened the temperatures in all compartments begin to decrease. The temperature in the IHX compartment peaks at 2.8 hours (10080 s). The temperature in the RTC compartment decreases at a faster rate than that in the IHX compartments due to the relatively higher gas temperature and larger surface to volume ratio in the RTC compartment as compared to the IHX compartment. Figure 4 : Zone Temperature Evolution At the junctions in the IHX compartment there is also a decrease in the pressure difference due to the hydrostatic head term. The pressure over a junction is given in ASTEC by. The mass density in the IHX compartment decreases which is as a result of the change in the air/helium fraction in the IHX compartment. Due to the difference in the hydrostatic head term due to gas composition difference between the environment and IHX the pressure difference across the IHX-Environment junctions becomes negative resulting in inflow through the leakage junction into the reactor building at which point radionuclide release through the leakages ends. This occurs at 35 minutes (2100 s) and 5.5 hours (19800 s) for the lower and upper leakage junctions respectively as seen in Figure 5. Two leakage junctions were chosen to simplify the modelling, one to represent a ground level leakage and the other an elevated leakage. The leakage areas were conservatively chosen with respect to the public dose, such that it was equivalent to a building leakage of 50% per day. Inflow of air through leakages results in a further driving force for flow through the stack. Session 2 : "ASTEC", paper 2.1 8/19

9 Figure 5: Pressure Difference at Junction for the First 8.2 hours Phase 3 (t > 91 hours) Post-Blowdown Flow from the RTC to the IHX compartment reverses to become slightly negative after blowdown is completed at 91 hours. This is due to a lower pressure in the RTC compartment than in the IHX compartment. The temperature in the RTC compartment decreases faster as it is cooling at a more rapid rate than in the IHX compartment and with the end of blowdown the pressure in the RTC compartment falls below that of the IHX compartment. The magnitude of the flow at this time is small, approaching the range of diffusive flow and thus caution should be exercised when observing these results due to possible numerical discrepancies Damper Closure Case The option of damper closure at the stack at 91 hours results in a rapid pressure build-up in the IHX compartment leading to the pressure difference across the upper leakage junctions changing from negative to positive, as seen in Figure 6. Flow reverses for the upper IHX leakage path to the environment (Figure 2) so that flow is now toward the environment again whilst the lower IHX leakage path to the environment (Figure 2) continues to intake air into the Reactor Building as seen in Figure 7. The difference in pressure behaviour at the junctions is due to the hydrostatic head term. The result is that a natural convection loop in the IHX renews release to the environment through the upper IHX leakage path. Figure 6: Pressure difference at the Junctions up to 277 hours in the Transient Session 2 : "ASTEC", paper 2.1 9/19

10 Figure 7: Mass flow at the Junction up to 277 hours in the Transient mm Break Results Results are presented below for the distribution of the dust in the reactor building for the case with no damper. Results are also presented for the I-131 as a gas and as and aerosol (not attached to dust) for the cases with and without damper. Note that the dust results for the case of damper closing have not been listed due to damper closing having an insignificant effect on the distribution of the dust since most dust deposits before damper closing. Therefore, the results in Table 2 can be considered as representative for both cases with/without the damper. Table 2: Dust Distribution (% of Total Dust Mass Injected) at 300 Hours (4 mm Break, No Damper) Dust Deposited on Dust Deposited Walls on Ceiling Dust Deposited on Floor Total Dust Deposited in Zone Dust Suspended in Zone RTC Compartment < IHX Compartment 0.05 < DVS Compartment < 0.01 < < 0.01 FIP Compartment < 0.01 < < 0.01 FHL1 Compartment < 0.01 < < 0.01 SIP Compartment < 0.01 < < 0.01 RRB Compartment < 0.01 < < 0.01 TOTAL RETENTION IN BUILDING ENVIRON1 (Stack release) < 0.01 < 0.01 < 0.01 < ENVIRON2 (Leakages at 20.0 m) < 0.01 < 0.01 < 0.01 < ENVIRON3 (Leakages at 0.7 m) < 0.01 < 0.01 < 0.01 < 0.01 < 0.01 TOTAL RELEASE TO ENVIRONMENT < 0.01 < 0.01 < 0.01 < NOTE: No Filtration was modelled. Session 2 : "ASTEC", paper /19

11 Table 3: I-131 Distribution (% of Total I-131 Mass Injected) at 300 Hours for 4 mm Break No Leakage, No Damper With Leakage, No Damper With Leakage, With Damper With Leakage, No Damper, Aerosol Sensitivity Case With Leakage, With Damper, Aerosol Sensitivity Case RTC IHX DVSW FIP FHL SIP RRB < 0.01 < < TOTAL RETENTION IN BUILDING ENVIRON1 (Stack Release) ENVIRON2 (Leakage) ENVIRON3 (Leakage) TOTAL RELEASED TO ENVIRONMENT < < < 0.01 < 0.01 < 0.01 < NOTE: No Filtration was modelled mm Break Study Phase 1 (t < 4 s) Rupture Panels Open High helium mass flow rates into the RTC compartment occur at the beginning of the break which results in the rupture panels all opening within the first 4 s as seen in Figure 8. Due to the high temperature of the incoming gas the temperature in the RTC compartment increases rapidly in the first few seconds peaking at a value just over 500 K after 11 s (see Figure 10). Session 2 : "ASTEC", paper /19

12 Figure 8: Zone Pressure during phase Phase 2 (t < 18 Hours) Completion of Blowdown After blowdown, the pressures of all compartments, including the IHX compartment, decrease, see Figure 9. At the junctions in the IHX compartment there is also a decrease in the pressure difference due to the term. The mass density in the IHX decreases as a result of the change in the air/helium fraction in the IHX compartment, see Figure 11. The helium mass fraction in the RTC compartment and other compartments increases with the RTC compartment approaching 100% He after approximately 25 seconds. The pressure difference across the lower and upper leakage junctions to the environment from the IHX becomes negative at s, as seen in Figure 12, resulting in inflow through that leakage junction into the reactor building at which point radionuclide release through the leakages ends. Inflow of air through leakages results in a further driving force for flow through the stack. Figure 9: Zone Pressure Evolution up to 76 seconds Session 2 : "ASTEC", paper /19

13 Figure 10 : Zone Temperature Evolution up to 76 seconds Figure 11: Helium Mass Fraction up to 76 seconds Figure 12 : Junction Pressure Difference up to 76 seconds Session 2 : "ASTEC", paper /19

14 The temperatures in the compartments drop at different rates, see Figure 13, depending on the temperature and the surface area to volume ratio. The cooling of vent shaft compartments causes the flow from the vent shaft to the environment to reverse at around 7 minutes, see Figure 15, and release of radionuclides and dust stops at this point. The building pressure begins to increase slightly due to inflow of air through stack and leakages, see Figure 14. At approximately 3.5 hours the flow through the vent shaft reverses and release through the stack occurs again, see Figure 16. Figure 13: Zone Temperature Evolution up to 16 minutes Figure 14 : Pressure Evolution up to 16 minutes Session 2 : "ASTEC", paper /19

15 Figure 15 : Junction Mass Flow up to 16 Minutes Figure 16: Junction Mass Flow up to 5.5 hours Phase 3 (t > 18 Hours) At approximately 18 hours, after blowdown is complete, the flow from the RTC compartment to IHX reverses. This has important implications for the release to environment of radionuclides released from the core during core heat-up as the major part is released after the mass flow between the RTC and IHX compartment reverses. Thus, radionuclides released into the RTC compartment after the flow reverses are mainly retained in the RTC compartment. Session 2 : "ASTEC", paper /19

16 Figure 17: Junction Mass Flow up to 22 hours Damper The option of damper closure at the stack at 50 hours results in a rapid pressure build-up in the IHX compartment leading to the pressure difference across the upper leakage junctions becoming positive from having been negative prior to damper closure, see Figure 18. Flow reverses for the upper IHX leakage path so that flow is now to the environment again whilst the lower IHX leakage path continues to intake air into the RBB as seen in Figure 19. The difference in pressure behaviour at the junctions is due to the term. The result is that a natural convection loop in the IHX renews release to the environment through the upper IHX leakage path. Figure 18 : Pressure Difference up to 83 Hours Session 2 : "ASTEC", paper /19

17 Figure 19: Junction Mass Flow up to 83 hours Results Table 4: Dust Distribution (% Total Dust Mass Injected) at 300 Hours for 100 mm Break (No Damper & Damper*) Dust Deposited on Walls Dust Deposited on Ceiling Dust Deposited on Floor Total Dust Deposited in Zone Dust Suspended in Zone RTC Compartment IHX Compartment <0.01 DVS Compartment <0.01 FIP Compartment < <0.01 FHL1 Compartment <0.01 < <0.01 SIP Compartment <0.01 < <0.01 RRB Compartment <0.01 <0.01 <0.01 <0.01 <0.01 TOTAL RETENTION IN BUILDING ENVIRON1 (Stack Release) <0.01 <0.01 <0.01 < ENVIRON2 (Leakages at 20.0 m) <0.01 <0.01 <0.01 <0.01 <0.01 ENVIRON3 (Leakages at 0.7 m) <0.01 <0.01 <0.01 <0.01 <0.01 TOTAL RELEASED TO ENVIRONMENT <0.01 <0.01 <0.01 < NOTE: No Filtration is modelled in the ASTEC calculation * - Damper and No Damper cases have similar results Session 2 : "ASTEC", paper /19

18 Table 5: I-131 Distribution (% of Total I-131 Mass Injected) at 300 Hours 100 mm Break No Leakage, No Damper With Leakage, No Damper With Leakage, With Damper With Leakage, No Damper, Aerosol Sensitivity Case RTC IHX DVSW FIP FHL SIP RRB < 0.01 Total Retention in Building ENVIRON1 (Stack Release) ENVIRON2 (Leakage) 0.00 < < 0.01 ENVIRON3 (Leakage) 0.00 < 0.01 < 0.01 < 0.01 Total Released to Environment NOTE: No Filtration is modelled in the ASTEC calculation 4. CONCLUSION An evaluation of the applicability of the ASTEC code to simulate HTR accidental scenarios has been carried out. Although initially developed in the field of LWRs ASTEC contains models which make them suitable for out of core analysis of HTRs. New models need to be developed for analysis of HTR cores whilst the main requirement for the out of core models is extensive verification and validation. An analysis of a reactor building performance following a 4mm and 100mm pipe break releasing helium, dust and radionuclides from the coolant system into the reactor building has been carried out. The results show that a vented reactor building provides significant retention of dust and radionuclides whilst relieving pressure build-up. These results are preliminary and, in order to improve confidence in them, it is proposed that future code comparison exercises and verification against experiments are performed. Analysis has been performed for blowdown and post-blowdown situations. REFERENCES [1] Electric Power Reseach Institute (EPRI), The Next Generation Nuclear Plant: The Case for Utility Interest, [2] Van Dorsselaere J.P., Seropian C., Chatelard P., Jacq F., Fleurot J., Giordano P., Reinke N., Schwinges B., Allelein H.J., Luther W., "The ASTEC integral code for severe accident simulation, Nuclear Technology, vol.165, March 2009 [3] Fontanet J. et al, PBMR Confinement Analysis During Helium Pressure Boundary Breaks, Proceedings HTR2008: 4 th International Topical Meeting on High Temperature Reactor Technology, Washington, D.C. (USA), September 28 October 1, Session 2 : "ASTEC", paper /19

19 [4] Allelein H.-J. et al, COCOSYS: Status of development and validation of the German containment code system, Nuclear Engineering and Design 2008, vol.238, N 4, pp [5] Gauntt R.O. et al, MELCOR Computer Code Manuals, Reference Manual, Version 1.8.5, Vol.2, Rev.2, Sandia National Laboratories, NUREGICR-6119, SAND /1. [6] Murata K.K. et al, User s Manual for CONTAIN 1.1: A Computer Code for Severe Nuclear Reactor Accident Containment Analysis, Sandia National Laboratories Report SAND , April [7] Van Der Merwe J.J.. et al., Validation and Verification of Flownex Nuclear, Proceedings HTR2006: 3 rd International Topical Meeting on High Temperature Reactor Technology, Johannesburg, South Africa, October 1-4, [8] Mazana N., A thermal hydraulic analysis of pipe breaks in the pebble bed modular reactor main power system, International Conference on Nuclear Engineering 16 (ICONE 16), Orlando, Florida (USA), May 11-15, [9] 10 CFR , Standards for Protection Against Radiation, Definitions, Code of Federal Regulations (CFR) Title 10, Energy, Part , December 19, 2002, [10] General Atomics, GT-MHR Preliminary Safety Assessment Report, DOE-GT-MHR , September [11] Van Dorsselaere J.P., Reinke N., Status and Perspectives of ASTEC Models, Proceedings ERMSAR 2008 (European Review Meeting on Severe Accident Research), Nesseber (Bulgaria), September 23-25, [12] Ball S.J. and Fisher S.E., Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs), Oak Ridge National Laboratory NUREG/CR-6944 (March 2008). [13] Murata K.K. et al, Code Manual for CONTAIN 2.0: A Computer Code for Nuclear Reactor Containment Analysis, NUREG/CR-6533, [14] Allelein H.-J., Neu K., Van Dorsselaere J.P., Müller K., Kostka P., Barnak M., Matejovic P., Bujan A., Slaby J., European validation of the integral code ASTEC (EVITA), Nuclear Engineering and Design, 221 (2003) [15] Van Dorsselaere J.P. and Allelein H.-J., ASTEC and SARNET Integrating Severe Accident Research in Europe, Proceedings EUROSAFE Forum 2004, Berlin (Germany), Nov [16] Albiol T. et al, Presentation of SARNET2, Proceedings ERMSAR 2008 (European Review Meeting on Severe Accident Research), Nesseber (Bulgaria), September 23-25, [17] Kljenak I., Dapper M., Dienstbier J., Herranz L.E., Koch M.K., Fontanet J., Thermalhydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code, Nuclear Engineering and Design, vol.240 (March 2010), [18] Association of German Engineers, AVR: Experimental High Temperature Reactor: 21 Years of Successful Operation for a Future Energy Technology, ISBN , [19] Beychok M. R., Fundamentals of Stack Gas Dispersion, 4 th Edition, Session 2 : "ASTEC", paper /19