Modelling of QUENCH-03 and QUENCH-06 Experiments Using RELAP/SCDAPSIM and ASTEC Codes

To prevent total meltdown of the uncovered and overheated core, the reflooding with water is a necessary accident management measure. Because these actions lead to the generation of hydrogen, which can cause further problems, the related phenomena are investigated performing experiments and computer simulations. In this paper, for the experiments of loss of coolant accidents, performed in Forschungszentrum Karlsruhe, QUENCH-03 and QUENCH-06 are modelled using RELAP5/SCDAPSIM and ASTEC codes. The performed benchmark allowed analysing different modelling features. The recommendations for the model development are presented

ICONE-22450 RELAP5-3D CODE APPLICATION FOR RBMK-1500 REACTOR CORE ANALYSIS

ABSTRACT The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed quite good mutual coincidence of the calculation results and good agreement with real plant data. INTRODUCTION RELAP5 code originally was designed for PWR and BWR type reactors to provide the US Government and industry with an analytical tool for the independent evaluation of reactor safety through mathematical simulation of transients and accidents. In this paper RELAP5-3D code was evaluated for its suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. Using RELAP5-3D code a successful best estimate RELAP5-3D model of Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data [1]. The two benchmark problem analyses, that were performed during the validation of the successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor and reported here are: feedwater flow perturbation and reactor power reduction transients. Both benchmarks were modeled using the RELAP5-3D code and the calculation results compared to the calculation results obtained using the STEPAN code, specially designed fo

Application of Best Estimate Approach for Modelling of QUENCH-03 and QUENCH-06 Experiments

One of the important severe accident management measures in the Light Water Reactors is water injection to the reactor core. The related phenomena are investigated by performing experiments and computer simulations. One of the most widely known is the QUENCH test-program. A number of analyses on QUENCH tests have also been performed by different computer codes for code validation and improvements. Unfortunately, any deterministic computer simulation is not free from the uncertainties. To receive the realistic calculation results, the best estimate computer codes should be used for the calculation with combination of uncertainty and sensitivity analysis of calculation results. In this article, the QUENCH-03 and QUENCH-06 experiments are modelled using ASTEC and RELAP/SCDAPSIM codes. For the uncertainty and sensitivity analysis, SUSA3.5 and SUNSET tools were used. The article demonstrates that applying the best estimate approach, it is possible to develop basic QUENCH input deck and to develop the two sets of input parameters, covering maximal and minimal ranges of uncertainties. These allow simulating different (but with the same nature) tests, receiving calculation results with the evaluated range of uncertainties

Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes

Deterministic Analysis of Natural Circulation Events at the Ignalina NPP

Eight main circulation pumps (MCPs) are employed for the cooling of water forced circulation through the RBMK-1500 reactor at the Ignalina nuclear power plant (NPP). These pumps are joined into groups of four pumps each (three for normal operation and one on standby). In the case of all MCPs trip, the reactor shutdown system is activated due to decrease of coolant flow rate. At the same time, after the pump trip, the coolant to the reactor fuel channels during the first few seconds is supplied by pump coastdown. Later, the reactor is cooled by natural circulation. The main question arises whether this coolant flow rate is sufficient to remove the decay heat from the reactor core. This paper presents the investigation of all MCPs trip events at the Ignalina NPP by employing best estimate code RELAP5 and methodology of uncertainty and sensitivity analysis

Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here

Safety of nuclear energy in Lithuania and evolution after 1990

Serija: Lietuvos mokslas. Kn. 61. ISSN 1392-4044. CA WOS DB nurodo kitokį šaltinio ISBN 978-9986-795-46-9Preliminary works of Ignalina NPP construction have been started in 1974, and the first unit went into service in December 31 of 1983. At the same time the second unit was under construction and began construction of the third unit. The second unit has been planned to begin to start operate in 1986, but because ofChernobyl accident the works on unit operation have been moved in 1987. This second unit started to operate in August 31 of 1987. At that time 60 % of the third unit have been constructed already, but soon construction has been closed down. In 1998 on demand of "green party" the construction of the third unit has been cancelled. Now because of political reasons the Ignalina NPP reactor is shutdown, and the second unit is planned to shutdown at the end of 2009. Hence, all three units of Ignalina NPP were or will be shutdown for the political reasons. Despite of political reasons, the safety of nuclear energy in Lithuania has been provided and its level rose all the time, both for operating, and for the shutdown reactors. Requirements for safety of the nuclear power plants depends on the saved up experience, from a level of technical development of a society which grows continually and depends on the point of view of the state government. In this paper the evolution of safety of nuclear energy in Lithuania is discussed, beginning from the construction of the Ignalina Nuclear Power Plant up to the present daysLietuvos energetikos instituta

Safety analysis of beyond design basis accidents in RBMK-1500 reactors

At present the design basis accidents for RBMK-1500 are rather thoroughly investigated. The performed analyses helped to develop and implement a number of safety modifications. Further plant safety enhancement requires developing emergency procedures that would enable beyond design basis accidents management by preventing core damage or mitigating consequences of severe accidents. This paper presents results of Ignalina NPP Level 1 and Level 2 probabilistic safety assessment and their use for development of beyond design basis accidents list. The most important phenomena for RBMK type reactor severe accident management are described. The paper also presents physical processes that occur in an overheated reactor core and discusses its cooling capabilities. The discussion also includes processes that occur in the reactor and primary circuit surrounding compartments, i.e. confinement. A number of recommendations on accident processes management are presentedLietuvos energetikos instituta

Reliability assessment method of district heating network

Straipsnyje pateikiama centralizuoto šilumos tiekimo sistemos patikimumo vertinimo metodika, jungianti termohidraulinį, stipruminį-tikimybinį ir matematinį modelius. Kuriant modelį buvo taikomi baigtinių elementų, pirmos eilės patikimumo bei atsako paviršiaus, gedimų medžių, Monte Karlo modeliavimo metodai. Kaip pasiūlytos metodikos pavyzdys yra pa-demonstruotas termofikacinio vandens tiekimo į Kauno miesto centrą patikimumo vertinimas. Vertinant patikimumą buvo išanalizuoti galimi keliai, kuriais miesto centrą pasiekia termofikacinis vanduo, ir identifikuotos galimos vamzdynų avarijų vietos. Nustatytuose mazguose atlikti tikimybiniai ir termohidrauliniai skaičiavimaiA reliability assessment method of district heating systems, which covers thermohydraulic, strength-probabilistic and mathematical models, is presented. Methods of finite elements, first row reliability and response surface, failure trees, and Monte Carlo simulation were used in creation of the model. The reliability assessment of thermal water (hot water) supply to the centre of the Kaunas city is suggested as an application example of the proposed method. Evaluating reliability possible ways were analysed how thermal water reaches the town centre, and possible places of pipeline accidents were identified. Probabilistic and thermohydraulic calculations were carried out in determined unitsLietuvos energetikos institutasVytauto Didžiojo universiteta