Passive Decay Heat Removal System Research Articles

A numerical investigation of three-dimensional flows in large volumes in the context of passive containment cooling in BWRs

The paper describes Computational Fluid Dynamics (CFD) calculations undertaken in support of analyses of three-dimensional flows that take place in the drywell volumes of advanced boiling water reactors with passive decay–heat removal systems. Data for comparison are taken from the 1/40th-scale European Simplified Boiling Water Reactor (ESBWR) mock-up facility PANDA under conditions of symmetric steam injection and asymmetric outflow. Steady-state simulations for pure steam conditions illustrate how the separate flow streams mix to ensure balanced outflow conditions to the condenser units. A transient calculation has also been performed to examine how air, assumed to be released from solution in the PANDA boiler, would ultimately accumulate in the separate condenser units. Results provide a possible explanation for the rundown in performance of one of the condensers, behaviour which was repeatedly observed in some of the earlier PANDA tests. The work also provides more general insights on how trace amounts of non-condensable gases may accumulate in passive cooling equipment.

Thermal-hydraulic analyses of transients in an actinide-burner reactor cooled by forced convection of lead–bismuth

The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

Passive System Reliability Analysis: A Study on the Isolation Condenser

This paper deals with the reliability assessment of passive systems that have been developed in recent years by suppliers, industries, utilities, and research organizations, aimed at plant safety improvement and substantial simplification in its implementation. The present study concerns the passive decay heat removal systems that use, for the most part, a condenser immersed in a cooling pool. The focus of the paper is a reliability study of the isolation condenser system foreseen for advanced boiling water reactors (BWRs) for the removal of the excess sensible and core decay heat from the BWR by natural circulation. Furthermore, an approach aimed at the thermal-hydraulic performance assessment (i.e., the natural circulation failure evaluation) from the probability point of view is given. The study is not plant-specific-related but pertains to the conceptual design of the foregoing system.

Model development for analysis of the Korea advanced liquid metal reactor

The SSC-K code is under development for analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) design adopting a pool-type reactor in Korea. The SSC-L code which was originally developed at Brookhaven National Laboratory for analysis of a loop-type liquid metal reactor, is its precursory code. The main reason for the development is that SSC-L cannot be applied directly to the KALIMER design because its application is limited to only a loop-type reactor. The SSC-K code represents the core with multiple coolant channels incorporated with a point kinetics model for calculation of the reactivity feedback. It can provide detailed one-dimensional thermal-hydraulic simulations not only for the primary and secondary sodium coolant circuits, but also the steam/water circuit of the balance-of-plant. This paper presents an overview of the recent developments on the physical models for SSC-K. Those developments are concerned with the two-dimensional hot pool model for analysis of the thermal stratification phenomena in the hot pool, the model for the passive decay heat removal system, the sodium boiling model in the core, and other physical models necessary for the KALIMER analysis. It also demonstrates the analysis results for the unprotected accidents like unprotected transient over power, unprotected loss of flow, and unprotected loss of heat sink postulated in the preliminary KALIMER design. The major focus of these analyses is made on confirmation of the inherent safety characteristics for the reactivity feedback in the core.

Design of LMR reactor structures in the vicinity of hot pool free surface regions subjected to moving elevated temperature cycles

In this paper, the structural damage on the reactor structures in the vicinity of the hot pool free surface regions are investigated to check design validity and to propose a modified reactor baffle (RB) design with a Y-piece type cylindrical thermal baffle (TB) structure. Detailed thermal and stress analyses for an envelope thermal transient load of the Passive Safety Decay Heat Removal System event are carried out for both the original design and the modified design. From the results of the analyses, the maximum stresses for the RB and the reactor vessel occur in the level of hot pool free surface regions after primary pump flow coastdown. The modified RB design with the TB structure leads to a significant reduction of the thermal stress, the inelastic creep ratcheting strain, and the creep-fatigue damage compared with those of the original design. Finally, it is verified that the proposed design modification is acceptable in the Korea Advanced Liquid Metal Reactor structural design for hot pool free surface regions.

Core power limits for a lead-bismuth natural circulation actinide burner reactor

The Idaho National Engineering and Environmental Laboratory and Massachusetts Institute of Technology are investigating the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The design being considered here is a pool type reactor that burns actinides and utilizes natural circulation of the primary coolant, a conventional steam power conversion cycle, and a passive decay heat removal system. Thermal-hydraulic evaluations of the actinide burner reactor were performed to determine allowable core power ratings that maintain cladding temperatures below corrosion-established temperature limits during normal operation and following a loss-of-feedwater transient. An economic evaluation was performed to optimize various design parameters by minimizing capital cost. The transient power limit was initially much more restrictive than the steady-state limit. However, enhancements to the reactor vessel auxiliary cooling system for transient decay heat removal resulted in an increased power limit of 1040 MWt, which was close to the steady-state limit. An economic evaluation was performed to estimate the capital cost of the reactor and its sensitivity to the transient power limit. For the 1040 MWt power level, the capital cost estimate was 49 mills per kWhe based on 1999 dollars.

Experimental data base for containment thermalhydraulic analysis

This paper describes the joint research project DABASCO which is supported by the European Community under a cost-shared contract and participated by nine European institutions. The main objective of the project is to provide a generic experimental data base for the development of physical models and correlations for containment thermalhydraulic analysis. The project consists of seven separate-effects experimental programs which deal with new innovative conceptual features, e.g. passive decay heat removal and spray systems. The results of the various stages of the test programs will be assessed by industrial partners in relation to their applicability to reactor conditions.

Comparisons between Multidimensional and Lumped-Parameter GOTHIC Containment Analyses with Data

Advanced nuclear reactor concepts heavily rely on the availability and efficiency of passive cooling systems. This especially holds for advanced containment designs with passive decay heat removal systems that function by natural phenomena. Also, the development of catalyst modules for hydrogen mitigation measures is based on natural basic principles leading to hydrogen reduction and additional atmospheric mixing.To prove the functionability and availability of passive systems and their respective components, demonstration experiments at different scales are mandatory. In addition, it is the general perception that many more improved computational tools are needed for this purpose, where present lumped-parameter analysis methods are insufficient to provide the necessary information about local details and spatial distributions. Therefore, the next step in the development of analytical/numerical models is the transition/extension from lumped-parameter to multidimensional models and containment analysis codes.Also, recent posttest lumped-parameter analyses of the Heiss Dampf Reaktor H2 distribution experiment E11.2 with preexisting atmospheric stratification show a number of deficiencies compared with the data, indicating a need for more detailed modeling.The GOTHIC thermal-hydraulic containment code provides this required extension of the lumped-parameter model by incorporating multidimensional submodels for selected nodes (subcompartments). Applications of both model types to simulate hydrogen dispersion experiments in the Battelle Model Containment (BMC) demonstrate the limitations of the traditional approach and the improvement achieved by the multidimensional simulation. The importance of thermal and hydrogen concentration stratifications, the interactions with structural heat conductors, and the requirements to set up a consistent model when coupling lumped-parameter and multidimensional representations are discussed.Several hydrogen-mixing experiments performed in the BMC more than a decade ago were simulated with multidimensional GOTHIC models.Three types of modeling concepts have been tested:1. lumped-parameter model2. each compartment modeled two-dimensionally with the intercompartment connections simulated as flow path junctions3. full three-dimensional nodalization of the BMC, intercompartment connections simulated as gaps.The results of these GOTHIC calculations are compared with the experimental data and demonstrate the improvements that can be achieved by performing multidimensional containment simulations.

Preliminary study on the application of the heat pipe to the passive decay heat removal system of the modular HTR

To investigate the applicability of heat pipe to the decay hat removal (DHR) system of the modular HTRs, preliminary study of the Heat Pipe DHR System was performed. The results show that the Heat Pipe DHR System is applicable to the modular HTRs and its heat removal capability is sufficient. Especially, by applying the variable conductance heat pipe, possibility of a fully passive DHR system with lower heat loss during normal operation is suggested. The experiments to obtain the fundamental characteristics data of the variable conductance heat pipe were carried out and the experimental results are also shown.

Influence of the position and number of decay heat exchangers on the thermal hydraulics of a slab test facility: A comparison of analytical and experimental data

To assess the capability of passive decay heat removal systems of an advanced pool-type liquid-metal cooled reactor, natural circulation experiments have been performed to investigate the in-vessel cooling modes caused by the position and number of decay heat exchangers in operation. The rather simple slab test facility AQUARIUS is equipped with an electrically heated core and decay heat exchangers. Four different arrangements of heat exchangers are under consideration to study the temperature distribution and the flow behavior in the apparatus during both symmetrical and asymmetrical heat removal. The experiments have been carried out in water under laminar flow conditions. Temperatures have been measured under quasi-steady-state conditions. The observed flow paths have been documented photographically. In case of asymmetrical heat removal, especially when a single heat exchanger is operated in one of both upper plena, the temperature distribution and the flow behavior are different from symmetrical cooling modes. A comparison of analytically predicted temperatures using the COMMIX-2(V) computer program with experimental data shows reasonably matches the findings. The results of the numerically determined velocity fields are in good agreement with the visual observations.

Large scale experiments with a 5 MW sodium/air heat exchanger for decay heat removal

Sodium experiments in the large scale test facility ILONA were performed to demonstrate proper operation of a passive decay heat removal system for LMFBRs based on pure natural convection flow. Temperature and flow distributions on the sodium and the air side of a 5 MW sodium/air heat exchanger in a natural draught stack were measured during steady state and transient operation in good agreement with calculations using a two dimensional computer code ATTICA/DIANA.

Investigation into the occurrence of backflow effects in a heated rectangular channel during natural circulation

To address the capability of passive decay heat removal systems of an advanced liquid-metal-cooled nuclear reactor, natural circulation experiments were performed to investigate the bouyancy-driven backflow phenomenon in the mixed convection flow of a cooling channel. In the AQUARIUS half-slab test facility, the core is represented by a vertical channel formed by two parallel electrically heated walls. At the top end, the channel is connected to an upper plenum where the fluid is cooled by an immersion cooler. Fluid velocities and temperatures were measured under quasi-steady-state laminar conditions. The experiments indicate that for low heat flux densities and large channel widths, cold water from the upper plenum flows down and penetrates the heated channel where the flow direction is upward. A comparison of analytical predictions using the COMMIX-2(V) three-dimensional computer code against experimental AQUARIUS data shows reasonable overall agreement. The numerical results reflect that backflow effects were not predicted for the investigated problem.

ＦＢＲ炉容器直接冷却崩壊熱除去系の適用性評価

Reactor Vessel Auxiliary Cooling System (RVACS) is a simple and passive decay heat removal system for FBRs. However the heat removable capability of this system is small compared to that of immersed type of decay heat exchangers. So, RVACS has limit in applicability up to about 400MWe pool type LMFBR.The objectives of this report are to confirm its heat removal capability, to estimate its applicability limit of this RVACS for FBRs and to evaluate the effects of design parameters. In this report, the plant transient analyses have been performed with thermal-hydraulic code CERES. The main results are as follows:(1) The RVACS with heat collector have large heat removal capability to reduce the peak temperature of primary Coolant by 20°C than that without heat collector.(2) The effects of stack height and level difference of overflow liner on the heat removal capability are not so large.(3) The RVACS with heat collector is applicable up to 500MWe pool type LMFBR.

Analysis of rvacs tests for commix validation constantine

Two RVACS tests, A and B, were analyzed to support the validation of COMMIX and the design of Advanced Liquid Metal Reactors (ALMRs). The RVACS test facility provides a scaled simulation of the decay heat removal paths of a pool ALMR during the operation of the passive decay heat removal systems with the reactor simulated with electrically heated rods. Test A was characterized by a power input of 50 kW, natural convection in the sodium pool, forced RVACS air circulation, sodium pool overflow, and a heat up period of 8 h. In test B the power was 64 kW and the sodium pool heatup led to overflow at ∼ h from the beginning of the transient. In test A, after 7.5 h, the system reached near steady state with a temperature difference between the bottom and top of the pool of 96°C. In both tests the vertical temperature gradient in the hot pool remained small throughout the transient while the cold pool exhibited a significant stratification. The sodium temperature variations in the radial and azimuthal directions were insignificant. The COMMIX predictions for the sodium pool temperatures and the air outlet temperatures were in good agreement with measurements.

An Assessment of the Core Degradation Frequency in a Typical Large LMFBR Design for Internal Accident Initiators—A Comparison with PWR Predictions

A comparative assessment of the core degradation frequency due to internal accident initiators between a typical large liquid-metal fast breeder reactor (LMFBR) design and pressurized water reactors (PWRs) has been performed. For the PWR system, existing analyses have been utilized. For the reference LMFBR, an extensive analysis has been performed for two accident initiators, i.e., loss of off-site power and loss of main feedwater. Based on this analysis an estimate of about1 X 10/sup -6//reactor X yr has been obtained for the core degradation frequency of the reference LMFBR. This estimate is significantly smaller than the PWR core degradation frequency ( about 6 X 10/sup -5//yr). A sensitivity analysis shows that the parameters having the largest impact on the unavailability of decay heat removal are (a) for the ''loss of off-site power'' initiator: human error and failure to restore off-site power, and (b) for the ''loss of main feedwater'' initiator: the leakage rates of the passive decay heat removal system and the adoption of the policy to repair the Na-NaK heat exchanger only during normal shutdowns. The results indicate that the LMFBR system has the potential of higher resistance than the PWR system to the accident initiators considered. The lowermore » core degradation frequency estimated for the LMFBR system is due to the presence of two redundant and diverse reactor shutdown systems, with a self-actuated feature included in one of them, the incorporation of a passive decay heat removal system, and the significantly lower sensitivity of the reference LMFBR to primary system pipe breaks.« less