Study On Count Rate Correction Method In Spent Fuel Detection

Though the nuclear power plant can generate electricity,it will produce amounts of spent fuel which contain uranium(U,95%),plutonium(Pu,1%),minor actinides(1%)and long lived fission products(3%)at the same time.The minor actinides(Np,Am and Cm)and long lived fission products pose an enormous threat to the environment.The separation and purification of U,Pu also plays an important part in the reuse of nuclear fuel.The contents of U,Np,Pu should be detected in real time during the process of separation and purification because the spent fuel is highly radioactive and toxic.A new type of graphite pre-diffraction Energy Dispersive X-ray Fluorescence Spectrometer(EDXRF),designed by our school and China Institute of Atomic Energy,was used to detect the contents of U,Np,Pu.The preliminary experiment results show the spectrometer of feasibility and effectiveness.However,the measurement results are incorrect when the concentrations of the samples increase.As a result,the measuring linear range is limited.The graphite pre-diffraction EDXRF is designed based on digital nuclear spectrometer,and the performance indicators of energy resolution and count rate are used to test the performance of the spectrometer.The energy resolution is affected by the intrinsic energy resolution of detector,pile-up pulse,ballistic deficit and noise,and the count rate relatives to pile-up pulse and count losses.The techniques of pile-up pulse separation and count rate correction are studied based on the National Natural Science Foundation of China,called study of the digital separating techniques of high rate nuclear pulse signal on high resolution(granted No.11475036).In view of expanding measuring linear range,the structure of the digital nuclear spectrometer and the algorithms for filtering and shaping are studied.In the study of digital nuclear spectrometer structure,some passive filtering and shaping circuits,including C-R circuit,R-C circuit and pole-zero cancellation circuit are introduced.The function of the circuits is discussed with simulated nuclear signal.A recursive model for Sallen-Key circuit is established by using numerical analysis method and the optimized parameters are also selected by testing its filtering performance on a noisy nuclear signal.The front-end circuitry consists of one-stage C-R circuit,linear amplifier and low-pass Sallen-Key filter.Combined with high-speed ADC,FPGA and MCU,the structure of digital nuclear spectrometer is set up.On the other hand,the algorithms of Gaussian filtering and trapezoidal pulse shaping are explored.The application of trapezoidal pulse shaping in pile-up pulse separation is also discussed.However,the baseline of the shaped pulse drafts under the condition of the input signal associated with noise.It is especially obvious in the process of consecutive pulses.In order to remove the baseline draft,the method of setting the first data of the input signal to 0 is proposed based on theoretical analysis.A fast-slow system in which the pulses are processed by a fast channel in parallel with a slow channel is presented.The fast channel(with narrow shaped pulse)is applied to obtain incoming count rate,preparing data for count losses correction in the slow channel,while the slow channel(with wide shaped pulse)aims to creating high resolution energy spectrum.The ratio of the total count rate in the fast channel and slow channel is developed to correct the count losses in the slow channel.This method is called count rate correction method.It is demonstrated to correct the count losses due to high concentrations of samples to increase count rate.In the verification experiments,the count rate correction method is verified on an EDXRF system,and the dead time correction method is used as a contrast.Finally,the method is realized in spent fuel measurement system to expand the measuring linear range.