Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients (original) (raw)
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Nuclear Engineering and Design, 2007
Kinetic simulations of fission product activity in primary circuits of a typical PWR under power transients, has been performed. A detailed two-stage model-based methodology has been developed and implemented in a computer coder FPCART which uses LEOPARD and ODMUG codes as subroutines. For normal constant power operation, results for over 39 fission products show that the activity due to fission products in fuel region of PWRs is dominated by 134 I which is followed by 134 Te and 133 I. The value of the total fission product activity in fuel region predicted by FPCART code has been found to agree with-in 0.36% range with the corresponding values found by using the ORIGEN-2.0 code. The predictions of FPCART code have also been found in good agreement with the corresponding values found in ANS-18.1 Standard as well as with some available power-plant operation data with 2.4% deviation in the value of specific activity of the dominating fission product 134 I. The saturation value of the fission product activity in coolant depends strongly on the fuel-clad gap escape rate coefficient (ε) and approaches a maximum value with increasing value of ε. During power transients, the FPCART predictions have been found in good agreement with the corresponding experimental measurements of 131 I specific activity for Beznau and Surry PWRs.
Sensitivity analysis of fission product activity in primary coolant of typical PWRs
Progress in Nuclear Energy, 2011
A model has been developed for static and dynamic activity analysis of the fission product activity (FPA) in primary coolant of typical pressurized water reactors (PWRs). It has been implemented in the FPCART based computer program FPCART-SA. For long steady power operation of reactor, the computed values of normalized static sensitivity have been compared with the corresponding values obtained by using the dynamic sensitivity analysis. The normalized sensitivity values for the reactor power (P), failed fuel fraction (D), coolant leakage rate (L), total mass of coolant (m) and the let-down flow rate (Q) have been calculated and the values: 1.0, 0.857, À2.0177 Â 10 À6 , 2.349 Â 10 À4 , À2.329 Â 10 À4 have been found correspondingly for Kr-88 with the dominant value of FPA as 0.273 mCi/g.
Effect of flow rate transients on fission product activity in primary coolant of PWRs
Progress in Nuclear Energy, 2007
Simulations of fission product activity in primary circuits of a typical PWR under flow rate transients have been performed by using a two-stage model for release of fission products from fuel into coolant region. A one-dimensional nodal-scheme has been developed for modeling the behavior of fission products in the primary circuit. For constant-power operation at constant flow rate, results for 15 major fission products show that the activity due to fission products in the primary coolant circuit of PWRs is dominated by 133Xe and it is followed by 135Xe, 131MXe and 129Te which contribute 40%, 12.9%, 11% and 8.2%, respectively, to the total fission product activity. The results of these simulations have been found to agree well with the corresponding values found in ANS-18.1 Standard as well as with some available power-plant operation data. These simulations indicate a strong dependence of saturation values of specific activity on primary coolant flow rate. For pump coastdown having a characteristic time tp ∼ 2000 h, a 8.6% increase has been observed in the value of total specific activity due to fission products. For increasing tp values, the value of maximum specific activity due to fission products shows a rise followed by an approach towards a saturation value.
Evaluation Of Radionuclides Release Estimation Of Power Reactor Using Scdap / Relap
KnE Energy, 2016
Incident of radiation release to the environment is important event in reactor safety analysis. Numerous studies have been conducted using various computer codes, including SCDAP/RELAP, to calculate radionuclide releases into the reactor coolant during severe accident. This paper contains description of calculation results of radionuclide release from reactor core to primary coolant system in a1000 MW PWR reactor with the aim to study behavior of radionuclide releases during severe accident. The calculations using SCDAP/RELAP was done by assuming that there has been a station black out which ends up with some vapor released into the containment. As a result, the water level in core was reduced up to a level where the core is no longer covered by water. The uncovered core heats up to certain temperature where the oxidation of the cladding started to occur. Afterwards the oxidation generated heat made fuel melting temperature reached and as consequences the release of radionuclide to...
Modelling of fission-product transport in the reactor coolant system
Annals of Nuclear Energy, 2013
The Phébus fission product (FP) programme studies the phenomenology of severe accidents in watercooled nuclear reactors. Five tests were performed in the frame of the programme covering fuel-rod degradation and FP behaviour released via the coolant system into the containment. To model FP transport and behaviour in the coolant system, numerous physical and chemical phenomena have to be taken into account. In the vapour phase, for example, FP speciation, vapour condensation and vapour/surface or vapour/aerosol reactions have to be considered. The aerosol phase has to be modelled with nucleation, growth and deposition processes. Finally, remobilisation phenomena like resuspension and revaporisation have to be taken into account for delayed release into the containment. Four Phébus FP tests (FTP0, FPT1, FPT2, FPT3) have been modelled with the ASTEC/SOPHAEROS code. Modelling shows an overall good estimation of retention for the main FPs (e.g., I, Cs, Mo). Furthermore, a strong connection is revealed in the gaseous phase chemistry between I, Cs, Cd and Mo which has a great impact on gaseous iodine release into the containment. The Phébus FP test modelling also exposes disagreement on FP retention when laminar gaseous flow is not well developed. Finally, probably the most significant shortcoming in modelling that Phébus-FP tests highlighted concerns vapour-phase iodine-chemistry modelling at low temperature. The study of this latter point is on going with the experimental programme ISTP/CHIP.
Nuclear Engineering and Design, 2021
In order to perform this comparison, some simple common criteria related to accidental behavior of the reactors have been developed. The first kind of criteria aims at assessing the main physical thresholds to exceed in order to have a core degradation: phase changes of coolant and of core materials (including the effect of chemical reactions) for the various reactor concepts considered. The second set of criteria deals with kinetics aspects of the accident. On the basis of core power (after scram and without scram), on the coolant inventory and on the reactor capability to be passively cooled, the heating rate of the coolant and of the core materials are assessed thanks to simplified energy balances presented in the paper. As a result, for each reactor concept, the time to reach the physical thresholds defined above is evaluated. A third set of criteria deals with core features and aims at assessing the possible reactivity insertion that withstands each concept up to core melting (or boiling for the MSR) and the associated expected power peaks in case of coolant voiding/depressurization and in case of fissile material compaction. Finally, a last criterion set deals with the assessment of the possibility to jeopardize physical barriers confining fission products. These criteria deal with the risk of barrier loadings due to coolant and core material vaporization depending on the features of the coolant/fuel and on the operating point of each reactor concept. In the last part of the paper, a synthesis is made in order to underline the weak and strong points of each of the reactor concepts investigated in terms of severe accident prevention and mitigation.
2009
The National Commission of Nuclear Energy (CNEN), which is the Brazilian nuclear regulatory commission, imposes safety and licensing standards in order to ensure that the nuclear power plants operate in a safe way. For licensing a nuclear reactor one of the demands of CNEN is the simulation of some accidents and thermalhydraulic transients considered as design base to verify the integrity of the plant when submitted to adverse conditions. The accidents that must be simulated are those that present large probability to occur or those that can cause more serious consequences. According t o the FSAR (Final Safety Analysis Report) the initiating event that can cause the largest damage in the core, of the IEA-R1 research reactor at IPEN-CNEN/SP, is the LOCA (Loss of Coolant Accident). The objective of this paper is estimate the frequency of the IEA-R1 core damage, caused by this initiating event. In this paper we analyze the accident evolution and performance of the systems which should mitigate this event: the Emergency Coolant Core System (ECCS) and the isolated pool system. They will be analyzed by means of the event tree. In this work the reliability of these systems are also quantified using the fault tree.
The Impact of Reactor Operating Time on Activation Levels for Safety Analyses
Fusion Technology, 1992
Radioactivity induced in a typical fusion power reactor was calculated for all elements with atomic number Z < 84 and for different irradiation times. It was shown that the shutdown activity varies widely with the duration of the irradiation time. In general, the activity induced by radionuclides with halflives that are significantly longer than the period of irradiation increases with increasing the irradiation time. On the other hand, the level of activity generated by any radionuclide with a half-life which is significantly shorter than the reactor lifetime reaches a peak early during irradiation and then may starts to drop to lower value before the end of irradiation. The severity of this peaking is determined by the destruction rate of the parent element The activities generated by long-lived nuclides (important for waste management) in any fusion reactor with life time in the order of 30 years reach their peak values at end-of-life. In the mean time, using the activity and decay heat values generated by short and intermediate-lived radionuclides at the end of reactor life to represent the worst case values used in safety analyses related to a loss of coolant accident (LOCA) and accidental release of radioactive inventory might lead to a substantial underestimation of the results.