Keywords

1 Introduction

The radioactivity of the gamma source is usually measured by the gamma spectrometer. The crystal types of the gamma detectors include sodium iodide (NaI), cadmium zinc telluride (CZT), and high-purity germanium (HPGe). Among which, the HPGe detector is of excellent energy resolution and widely used in the field of radioactive environment monitoring, nuclear installation effluent monitoring, and radioactive waste measurement.

The principle of the HPGe gamma spectrometer for measuring the gamma radioactivity is through analyzing the measured gamma spectrum of the radioactive source, and the radioactive nuclides are identified by matching the full-energy peaks with the characteristic peaks in the library. The net counts and the peak radioactivity under the full-energy peak are calculated, as well as the nuclide radioactivity. Before the gamma spectrum is analyzed, the calibrations for the energy response, full width at half maximum (FWHM), and detective efficiency must be performed. Among those, the efficiency calibration is the most difficult part.

The main characteristics of the complex gamma radiation field are many radioactive nuclides inside, high background level, complex geometrical structure, sampling difficulty, narrow space nearby, etc. Here are the reasons why the HPGe gamma spectrometer system used in the conventional radiation laboratory is usually unsuitable for measuring the radioactivity of the complex gamma radiation field: (1) The majority of influence imposed by the around background level and other uninterested radioactive sources cannot be shielded because the conventional system is not equipped with the collimator and other specific measurement equipment. (2) It is difficult to acquire the detector efficiency by means of relative efficiency calibration method because it is very difficult to prepare the calibration-purpose standard source which shape is similar to that of the measured object. (3) It is difficult to obtain the specimen because the measured gamma source is usually the tightly sealed solid source.

Nowadays, the improved systems (such as ISOCART and ISOCS) have been developed for measuring the gamma radioactivity in complex radiation field. The collimator, cart, and “efficiency calibration software dispensing with standard sources” (such as GammaCalib software) have been equipped for those systems. The above configuration overcomes parts of the shortcomings of the conventional system, but it is still necessary to make several improvements: (1) The applicable range of the recommended HPGe detector is dissatisfactory because it is unsuitable either for the higher counting rate or for the higher gamma energy (E >2 MeV). (2) The system is difficult to move to the measurement location because the cart has a bigger volume and the collimator is very heavy. (3) The customization procedure of “the efficiency calibration software dispensing with standard sources” is complicated: (A) It is necessary not only to perform the detector efficiency calibration by using the standard sources in the characterization experiment, but also to calculate the detector efficiency by using MCNP program. (B) It is necessary to perform the theoretical characterization calculation of the detector efficiency for a great deal of photons with different incident angles on a large-scale parallel computation system. (4) The detector performance degeneration with time is not considered, so there has the risk that the error of measured radioactivity is bigger.

Compared with the above-mentioned system, the system described in this paper has the following advantages: (1) The HPGe detector is suitable for the bigger radiation range and gamma energy rang because the detector has been customized and optimized. (2) It is more convenient to move the system to the measurement location because of the smaller weight and volume of the collimator and the cart trestle. (3) The configuration of the cart trestle permits to more conveniently adjust the measurement height and orientation. (4) The calibration method is more convenient. The measurement precision and the modeling and computation time of the system described in this paper is equivalent to that of the above-mentioned system.

2 Radioactivity Measurement System

The main components of the gamma-radiation measurement system include: the hardware and software of ISOCART system, GammaCalib software, MERCURE program, MICROSHIELD software, and a set of standard sources used for detector calibration.

2.1 Main Instrumentation and Equipment

The hardware of ISOCART system include the HPGe detector (GEM10P4), the portable digital spectrometer (digiDART [1]), the positive high bias voltage/intelligent detector connection device (SMART-1-P), the collimator, 3-liter dewar (CFG-PG4-3), and the cart (ISOCART).

Choosing what kind of HPGe detector is mainly depended on the magnitude of the measured nuclide radioactivity and the gamma-ray energy range. For the high count rate measurements, the LEGe detector had better be chosen. For low radioactivity measurements, the high-efficiency coaxial germanium detector is more suitable. GEM10P4 is a p-type coaxial HPGe detector with wide energy response range (form 40 to 10 meV), 10% detector efficiency, high energy resolution (1.75 keV at 1.33 meV; 0.80 keV at 122 keV), and 41:1 Compton-peak ratio. Those characteristics permit that it is suitable for in situ measurement in the complex gamma-radiation environment.

DigiDART integrates the functions of the digital spectrometer, keyboard, screen display, processing software and storage, can independently accomplish the calibration, parameter setting, measurement, and analysis, and can transfer data with a computer via USB port.

SMART-1-P adopts the integrated front amplifier, high bias voltage, and high voltage protection signal interface, greatly simplifies the cable arrangement, and is especially suitable for in situ measurement and mobile measurement.

2.2 Computer Software

2.2.1 GammaVision Software

GammaVision software [2] is the main support software of ISOCART system. This software integrates the functions of the gamma-spectrum acquirement, intelligent multi-channel analyzer (MCA), radioactivity quantitative analysis, and microcomputer workstation. By using this software, it is easy to process the gamma-spectrum file, obtain the gamma spectrum, perform the energy and efficiency calibration, mark out the regions of interest (ROI), compute the nuclide radioactivity, edit the database, and print the measurement reports, etc.

2.2.2 GammaCalib Software

GammaCalib software [3, 4] which was developed by Beijing Institute of Applied Physics and Computational Mathematics and obtained state-level appraisal is a type of “gamma-detector efficiency calibration software dispensing with standard sources.” This software adopts the numerical integration method to calculate the detector efficiency, in-builds the nuclide (H to Pu) macroscopic cross sections interacted with the different-energy photons. Before delivered to the user, this software needs to be customized. The detector characterization is the most important step in the customization procedure. The main purpose of the characterization is to obtain the detailed geometrical parameters and accurate efficiency of the detector for all photons with the different incidence angles and energy. With the aid of the powerful functions of UNIGRAPH software, three-dimensional (3D) visualization modeling for the source and shields with arbitrary shapes can be realized by using GammaCalib software.

2.2.3 MERCURE Program

MERCURE program [5] developed by France CEA is a 3D photon/neutron transport calculation software using the Monte Carlo method. This software solves the integral-form particle transport equation: under the multi-group approximation, the Monte Carlo method is adopted to perform the integration calculations for the attenuation of the point kernels in straight lines.

2.2.4 MICROSHIELD Software

MICROSHIELD software is a simple and easy-to-use photon/neutron shielding calculation software using point-kernel integration method.

3 Radioactivity Measurement Method

3.1 Absolute Efficiency Calibration Method

In absolute efficiency calibration method [6], it is assumed that the detector efficiency is up to the following two independent coefficients:

  • Absolute efficiency K(E) which is only related to the characteristics of the detector such as the geometrical structure, dimension, and material composition.

  • Gamma-flux transfer function \( \varphi (E) \) which is only related to the characteristics of the shields and the gamma source.

K(E) is determined by the detector efficiency calibration experiment:

$$ K(E) \equiv \frac{{{\text{detected}}\;{\text{photons}}\;{\text{per}}\;{\text{second}}}}{{{\text{photon}}\;{\text{flux}}\;{\text{at}}\;{\text{crystal}}\;{\text{centre}}}} = \frac{S(E)}{{T_{a} \times \phi (E)}} $$
(1)

Here, S(E) is the net count under the full-energy peak; \( T_{a} \) is the alive time (second); \( \phi (E) \) is the gamma flux at crystal center imposed by standard sources (Bq/cm2).

3.2 Radioactivity Calculation Method

After the in situ gamma spectrum has been measured, the radioactivity or specific radioactivity A is calculated through the equation below:

$$ A = \frac{{\phi^{*} (E)}}{\varphi (E)} = \frac{S(E)}{{T_{a} \times K(E) \times\Gamma (E) \times \varphi (E)}} $$
(2)

Here, the dimension of A is Bq, Bq/cm2, or Bq/cm3; \( \phi^{*} (E) \) is the gamma flux at the crystal center (Bq/cm2); \( \varphi (E) \) is the gamma-flux transfer function (1/cm2 or cm); \( \varGamma (E) \) is the gamma-ray branching ratio.

MERCURE program is used to calculate \( \varphi (E) \). In the calculation, the interested gamma energy range (0–2500 keV) is divided into 10 groups; the gamma flux at the crystal center imposed by linear gamma ray is calculated for one unit radioactivity of the source with specific shape (point, surface, or volume). The calculated gamma flux is equal to the gamma-flux transfer function in numerical value. As other Monte Carlo calculations, the necessary data such as the geometrical structure and material composition of the gamma source and the shields (including the detector and collimator) should be collected.

The most interested part is to calculate the radioactivity or specific radioactivity \( A_{D} \) of the deposits on the inner surface of the pipe:

$$ A_{D} = \frac{{\varphi_{T} (E) - \varphi_{W} (E) \times A_{W} \times\Gamma (E)}}{{\varphi_{D} (E) \times\Gamma (E)}} $$
(3)

Here, \( A_{D} \) and \( A_{W} \) (measured by sampling method) are the radioactivity or specific radioactivity of the deposits and the water, respectively; \( \phi_{T} (E) \) is the gamma flux at the crystal center (Bq/cm2); \( \varphi_{D} (E) \) and \( \varphi_{W} (E) \) are the gamma-flux transfer function of the gamma sources in the deposits and water, respectively.

In order to facilitate the calculation of the flux transfer function, CALTF program has been developed by using the spline interpolation method. The database of the flux transfer function necessary for CALTF program is prepared for various pipe diameters, thickness, and length (the calculation matrix).

3.3 Radioactivity Modification Method

The calculated radioactivity value unavoidably has some errors, because the input data and the calculation models have some errors. In order to reduce the errors, considering that the dose rate can be easily and precisely measured, through modifying the calculated radioactivity value by multiplying the ratio (radioactivity modification factor) of the measured dose rate to the calculated dose rate, we obtain the more precise measured radioactivity value.

The relationship between the dose rate and the radioactivity is as follows:

$$ DR = \sum\limits_{i = 1}^{I} {\left[ {A(E_{i} ) \times \psi_{DR} (E_{i} )} \right]} $$
(4)

Here, DR is the dose rate; A is the radioactivity; \( \psi_{DR} \) is the flux transfer function; E i is the energy under the full-energy peak i; I is the number of the full-energy peaks.

Under the same measurement configuration (geometrical structure, dimension and material constituent, etc.), the dose rate transfer function keeps constant. Assume that the proportional relation among the radioactivity under the full-energy peaks remains constant, so the following correlation can be used to estimate the measured radioactivity value:

$$ A_{{{\text{meas}} .}} \approx A_{{{\text{calc}}.}} \left( {\frac{{DR_{{{\text{meas}} .}} }}{{DR_{{{\text{calc}} .}} }}} \right) $$
(5)

Here, \( A_{{{\text{calc}} .}} \) is the calculated nuclide radioactivity; \( DR_{{{\text{meas}} .}} \) is the measured dose rate; \( DR_{{{\text{calc}} .}} \) is the calculated dose rate; \( DR_{{{\text{meas}} .}} /DR_{{{\text{calc}} .}} \) is the radioactivity modification factor.

It could be expected that the measured radioactivity basically matches with the measured dose rate because of the above-mentioned modification.

4 Radioactivity Measurement of Waste Drums

4.1 Introduction of the Basic Conditions

At present, a large number of radioactive waste drums are stored in the QS and QT buildings of Daya Bay Nuclear Power Plant. It is necessary to transport these drums in batches to Beilong low and medium radiation waste repository for long-term storage. According to the requirements of relevant state regulations and policies, before these drums are shipped out, the plant shall provide the relevant data such as the species, composition, content, radioactivity of the radioactive substances inside the drums, and the dose rate around the drums.

In order to verify the applicability and validity of the technology described in this paper, and to prepare for measuring the radioactivity of a large number of drums in the future, after consulting with the relevant departments of Daya Bay Nuclear Power Plant, we measured four representative drums in the QT building. The following table gives the parameters of these drums (Table 1).

Table 1 Waste drum parameters
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4.2 Detector Calibration Experiment

The detector calibration experiment is performed in the radiation laboratory (a class-2 metering station) in Daya Bay Nuclear Power Plant. Because it is estimated that the corrosion-product nuclides (such as Co-60, Ag-110 m and Mn-54, with long half-life) in the waste drums occupy the main contribution to the surrounding dose rate, Ba-133, Cs-137, Co-60, Y-88 standard sources are selected in the experiment in order that the gamma energy of the standard sources can cover the range of 0–2000 keV.

Before the experiment, the standard sources are bandaged together, and the collimator is removed from the cart to let the detector bare. In the Experiment, the standard source package directly faces the detector and is placed in the position of 30 cm away from the detector. The gamma-spectrum measurement and storage are performed by using digiDART. In order to evaluate the influence of the distance on the measurement results [such as K(E)], the gamma spectrum of 50 cm distance is also measured. For the sake of obtaining sufficient counts and reducing the influence of background level, 5-h measuring time is necessary.

After the experiment, the GammaVision software is adopted to perform the gamma-spectrum processing, mark out the interested regions (ROI), perform the energy and FWHM calibration, and calculate the net counts S(E) under the full-energy peak. The detector absolute efficiency K(E) is then calculated by using Eq. (1). The calculation results are shown in the Fig. 1.

Fig. 1
figure 1

The calibration curve of the detector absolute efficiency

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4.3 In Situ Gamma-Spectrum Measurement

Eleven measurement schemes are performed. The detector direction should be adjusted to directly face the axis of the waste barrel. In order to evaluate the influence of the relevant parameters on the measurement results, the detector is allocated in different direction, distance, and height. The main measured parameters include the collimator height, the distance between the collimator and the barrel, gamma spectrum, dose rate around the barrel surface, and dose rate at the detector center. The measured parameters are summarized in the Table 2.

Table 2 Summary of the measured parameters
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4.4 Efficiency Calibration Dispensing with Standard Sources

The measurement configuration analysis model includes the information related to the materials of the detector, collimator, drum and waste (the gamma source), the distance among the above objects, as well as the geometrical dimension of the above objects. Before this model is established by using GammaCalib, MERCURE, and MICROSHIELD, it is necessary to estimate the average density of each mixture material (such as the concrete and the waste) and the uniformed density of each constituent.

The estimation procedures for the uniformed density of each constituent in various mixture materials are as follows: (1) to estimate the mass percentage (mixture ratio); (2) to estimate the mass; (3) to estimate the volume and the volume ratio; (4) to estimate the uniformed density. The calculation results are shown in Table 3. The measurement configuration analysis model is shown in Fig. 2. The detector efficiency curve of the measurement scheme 1 is shown in Fig. 3.

Table 3 The uniformed density of each constituent in various mixture materials
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Fig. 2
figure 2

Measurement configuration analysis model

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Fig. 3
figure 3

Detector efficiency curve of the measurement scheme 1(1060142 A)

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4.5 Calculation of the Gamma-Flux Transfer Function

The gamma-flux transfer function and the dose rate transfer function are calculated for specific measurement configuration by using MERCURE program and MICROSHIELD software, respectively. The calculation methods are described in Sect. 3.1. The data used in the calculations are similar to those of GammaCalib software. The gamma-flux transfer function of the measurement scheme 1 is shown in the following figure (Fig. 4 and Table 4).

Fig. 4
figure 4

Gamma-flux transfer function of the measurement scheme 1(1060142-A)

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Table 4 Radioactivity measurement results
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4.6 Analysis of the Measurement Results

4.6.1 Measurement Results

The radioactivity measurement results of the waste drums calculated by “the efficiency calibration method dispensing with standard sources” are listed in the following table. In this table, the ratios of the radioactivity calculated by the absolute efficient calibration method to those calculated by “the efficiency calibration method dispensing with standard sources” are given as well.

As we can see, there exist several nuclides with longer half-life and bigger radioactivity in four types of waste drums. Among which, Co-60 (T 1/2 = 5.27 years), Ag-110 m (T 1/2 = 250 days), and Mn-54 (T 1/2 = 313 days) still exist in the concrete wastes drums, and Co-60, Ag-108 m (T 1/2 = 127 years), and Cs-137 (T 1/2 = 30 years) still exist in the metal waste drum. In addition, no other nuclides are measured because their half-life is relatively shorter; thus, the radioactivity have been attenuated to the negligible degree.

We can also see that the deviations between the radioactivities calculated by the two methods are in the range of ±3%. This result indicates that the calculation precision of the two methods is equivalent.

4.6.2 Optimized Measurement Conditions

By analyzing the influence of the measured parameters (such as the distance, height, direction, and time) on the radioactivity measurement results, we can obtain the suitable optimized measurement conditions which will be used in lots of measurement activities in the future.

The measured radioactivity in scheme 1, 2, 7, and 8 has a little difference. However, the measurement distance of those schemes is greatly various. That factor indicates that the influence of the measurement distance on the measurement results can be ignored. It is also worth noting that the measurement distance should be appropriate: shorter distance will induce higher count rate and more dead time. In serious conditions, the count rate is too high to perform the measurement (the signals are saturated). Longer distance will cause smaller count rate and longer measuring time. Measurement experience suggests that the dose rate in the detector center be within 2–5 mSv/h.

The measured radioactivity in schemes 3 and 4 has a little difference. However, the detector height of those schemes is greatly various. The above factor indicates that the influence of the detector height on the measurement results can be ignored. In addition, the detector height had better remain in the half-height of the waste drum, and it is necessary to ensure that the detector can “watch” the whole drum.

The measurement results show that: (1) the dose rate around the surface of C1 concrete barrel is much more uniformed; (2) however, the dose rate around the surface of C4 metal barrel is very uneven. The main reason is that: (1) before consolidation, the waste resin and the concentrated salt have been fully mixed in C1 barrel; (2) the skew position of the filter cartridge placed in C4 barrel causes the shielding-layer thickness in each orientation is different. In order to accurately measure the radioactivity inside C1 and C4 barrels, the dose rate at four azimuths should be measured, and the shielding-layer thickness around the barrel wall should be modified in the transfer function calculation.

In theory, the longer the measurement time, the better the measurement results. However, the radioactivity measurement is constrained by the funds, job schedule, onsite radiation conditions, etc. The gamma-radiation level around the measured waste barrel is high. Measurement experience suggests that when the measurement time is longer than 20 min, the shape of the full-energy peaks of the main interested nuclides in the gamma spectrum is satisfactory, and the effective counts in each channel are much higher than the background counts.

5 Conclusion and Prospection

Through the customization of HPGe detector, the structure optimization of the collimator and the cart trestle, the design of the calibration device, as well as the establishment of a whole set of the calculation and analysis method and related software, the Gamma-Radioactivity Nondestructive Measurement System for Complex Radiation Field has been developed. The nuclide radioactivity measurement results of four types of waste barrels show that the measurement precision of the absolute calibration method adopted by this paper is equivalent to that of “the efficiency calibration method dispensing with standard sources.”

Along with the development of China’s nuclear industry and nuclear technology application, and the demands for safer, friendlier environment, this system will have a wider range of purposes. In order to facilitate the practical application, the system needs some further improvements in the experiment verification, measurement precision, miniaturization, portability, automation, etc. This can be achieved by the corporation with the gamma spectrometer manufacturers at home and abroad and the related research institutes and colleges.