Accurate determination of technetium-99 (Tc-99) is very important because any overestimation will cause the examined radioactive wastes to be categorized into super C class, which dramatically increases the cost of waste management Herein, we demonstrated that by adopting the analytical method comprising TEVA resin pretreatment and ICP-MS measurement, the determined Tc-99 concentrations in representative waste stream samples from the Lan-Yu low-level radioactive waste temporary storage site in Taiwan were approximately two orders of magnitude lower than those determined from the beta radiation measurement using a low background liquid scintillation counter Two important concerns emerged from our results First, severe interferences from other nuclides residing in the matrix considerably affect the determination of Tc-99, even when a low background liquid scintillation counter was used Second, the currently used Tc-99/Cs-137 scaling factor should be carefully revised, or it might lead to a considerable overestimation of the Tc-99 concentration

Accurate technetium-99 determination using the combination of TEVA resin pretreatment and ICP-MS measurement and its influence on the Tc-99/Cs-137 scaling factor calculation

In comparison with 
			emerging power systems, the Pressurized Water Reactors (PWRs) have 
			many times higher post shutdown radiation levels, originating partly 
			from the fission products released to the primary coolant from 
			defective fuel pins in the core. This results in prolonging the 
			maintenance schedule and translates into substantial economic costs. 
			To minimize the plant maintenance time and to reduce the radiation 
			dose for plant operators and general public, a detailed knowledge of 
			radioactivity buildup and its kinetics is essential.
			In this work, a detailed methodology has been developed for modeling 
			and simulation of kinetics of fission product activity in primary 
			coolant loops of typical PWRs under steady-state and power 
			transients. For this purpose, a detailed three-stage methodology has 
			been developed and implemented in the computer code FPCART,which 
			uses LEOPARD and ODMUG codes as its subroutines. It has been coded 
			in Fortran-77 and uses adaptive Runge-Kutta-Fehlberg algorithm as 
			its base ODE- solver.Mathematical model is based on a coupled system 
			of first order, ordinary differential equations governing the 
			kinetics of dominant fission products within the fuel, fuel-clad 
			gap, and the primary coolant loops. Code is capable of handling 
			power transients, and takes into account the effects of purification 
			system as well. Simulation of fission product activity in primary 
			coolant under flow-rate transients have also been performed by using 
			a two-stage model from fuel to fuel-clad gap and then from gap to 
			primary coolant region. A one-dimensional nodal-scheme has been 
			developed for modeling the behavior of fission products in the 
			primary circuit.
			For normal constant power operation, results of over 39 fission 
			products show that activity due to fission products in the fuel 
			region of PWRs is dominated by 134I and is followed by 134Te and 
			133I. The value of the 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 for primary coolant activity have 
			been found in good agreement with corresponding values of ANS-18.1 
			Standard as well as with some power plant measured data with 2.4% 
			deviation in the value of specific activity of the dominating 
			fission product 134I. Similarly,for constant power operation and 
			constant flow rate, results for 15 major fission products show that 
			the activity 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. These simulations indicate a strong dependence of 
			saturation values of specific activity on primary coolant flow rate. 
			For pump coast-down having a characteristic time tp ~ 2000 h, an 
			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. 
			The simulation of primary coolant activity due to 85Kr, 87Kr and 
			135Xe chains, have been carried out using classic Runge-Kutta (RK4), 
			adaptive Runge-Kutta-Fehlberg (RKF), Adams-Bashforth-Moulton (ABM) 
			and Semi-Implicit-Extrapolation (SIA), with later two as stiff 
			solvers. Deviations were observed between the corresponding 
			predictions between the lumped and un-lumped systems, especially, 
			during the initial phase of the simulations.
			Finally, a stochastic model has been developed for simulation of 
			fuel failure time sequences by sampling time dependant intensity 
			functions. Then the three stage model based deterministic 
			methodology of FPCART code has been extended to FPCART-ST, which 
			simulate the random fuel failure sequences followed by burst release 
			of radioactive contents present in fuel-clad gap at that time, into 
			primary coolant coupled with power transients. The value of the 131I 
			activity in primary coolant predicted by FPCART-ST code has been 
			found in good agreement with the corresponding values measured at 
			ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with 
			constant release option have also been found in good agreement with 
			corresponding experimental values for time dependant 135I, 135Xe and 
			89Kr concentrations in primary coolant measured during EDITHMOX-1 
			experiments.

Kinetic and Stochastic Modeling and Simulation of Fission Product Activity in Primary Loop of a Typical PWR

Saeed Ehsan Awan, PhD, Department of Physics & Applied Mathematics, PIEAS, June 2012. "Kinetic Simulation, Sensitivity Analysis of Fission Product Activity and Source Term Evaluation for Typical Accident Scenarios in Nuclear Reactors” Supervisor Dr.Nasir. M. Mirza; Co-Supervisor: Dr. Sikander M. Mirza Department of Physics & Applied Mathematics, PIEAS, Nilore 45650, Islamabad.With growing demands of safe and reliable energy resources worldwide, nuclear power plants present viable option.A two third majority of these plants are PWRs. In comparison with their competitors, PWRs suffer from significantly higher dose rate due to radioactivity in the primary circuit which is dominantly contributed by corrosion and followed by fission products leakage from fuel.There has been extensive investigation in developing corrosion resistance alloys.But the problem of corrosion product activity in primary circuit has aggravated in view of trend towards high burn-ups, high temperatures, and longer-life time reactors.Under this scenario, the significance of fission products releases becomes even higher.The fission product activity (FPA) is considered to be the second leading contributor towards prevalent radiation levels in Pressurized Water Reactors (PWRs).The elevated radiation level results in delay and prolongation of routine repair/maintenance tasks of reactor’s cooling system, which not only reduces its effectiveness but also results in several million dollars revenue loss per power plant annually. However the reliable estimates of fission product activity (FPA) are also significant for the evaluation of fuel performance, assessment of radiological consequences in case of any accident releasing radioactivity and scheduling repair/maintenance tasks.The detailed knowledge about radioactivity build up and sensitivity analysis of fission product activity (FPA) is essential for reducing the plant maintenance time, which also helps to reduce the dose for plant operators and general public.In this work, first a model is developed for dynamic and static sensitivity analysis of fission product activity in primary coolant of typical pressurized water reactor (PWR). It has been implemented in the FPCART based computer program FPCART-SA that carries out sensitivity analysis of fission product activity (FPA) using both static as well as dynamic approaches.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, -2.0177 ~ 10-6, 2.349 ~ 10-4, -2.329 ~ 10-4 have been found correspondingly for Kr-88 with the dominant values of fission product activity (FPA) as 0.273 ƒECi/g.In the second part of this study, evaluation of time dependence of source term has been carried out for a typical reactor system.The modeling and simulation of release of radioactivity has been carried out by developing a computer program FPARA which uses the ORIGEN2 code as subroutine, for core inventory calculations.Time dependent release of fission product activity to the containment and air has been simulated for loss of coolant accident scenarios.For noble gases, iodine and for aerosols, the release rate studies have been carried out for different leakage rates from containment. Effects of fraction of source in the coolant that is directly available after the accident on volumetric fission product activity were studied.Results show that volumetric activity in the containment air for different fission products remains strong function of decay constants, leakage rates, retention factors, deposition rates and fractional release rates.

Kinetic simulation, sensitivity analysis of fission product activity and source term evaluation for typical accident scenarios in nuclear reactors

A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglasmore » Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.« less

/pdf/a-model-for-predicting-fission-product-activities-in-reactor-1dlvu5e0my.pdf

A. Husain

Papers

A model for predicting fission product activities in reactor coolant: application of model for estimating i-129 levels in radioactive waste