Neutron Coincidence Counting Research Articles

A processing circuit for overlapped pulse signals for a thermal neutron coincidence counter

We have developed a processing circuit for overlapped pulse signals. The overlapped pulse signals are generated when the pulse signals of a He-3 detector, by using a neutron coincidence counter, are connected to shift register coincidence electronics by an OR gate device. The developed circuit detects the overlapped pulse signals from among four input signals and produces new 50 ns pulse wide signals. We considered a case where two pulse signals are simultaneously overlapped among four signals. This circuit was tested with an ACP safeguards neutron counter (ASNC) for an advanced spent fuel conditioning process (ACP) and a 252Cf neutron source at high rates. The loss rate of the output signal was reduced by 1.27% for singles and 4.75% for doubles when compared with the OR gate device. Also the variation for the triples was much bigger.

Neutron coincidence counting with digital signal processing

Neutron coincidence counting is a widely adopted nondestructive assay (NDA) technique used in nuclear safeguards to measure the mass of nuclear material in samples. Nowadays, most neutron-counting systems are based on the original-shift-register technology, like the (ordinary or multiplicity) Shift-Register Analyser. The analogue signal from the He-3 tubes is processed by an amplifier/single channel analyser (SCA) producing a train of TTL pulses that are fed into an electronic unit that performs the time- correlation analysis. Following the suggestion of the main inspection authorities (IAEA, Euratom and the French Ministry of Industry), several research laboratories have started to study and develop prototypes of neutron-counting systems with PC-based processing. Collaboration in this field among JRC, IRSN and LANL has been established within the framework of the ESARDA-NDA working group. Joint testing campaigns have been performed in the JRC PERLA laboratory, using different equipment provided by the three partners. One area of development is the use of high-speed PCs and pulse acquisition electronics that provide a time stamp (LIST-Mode Acquisition) for every digital pulse. The time stamp data can be processed directly during acquisition or saved on a hard disk. The latter method has the advantage that measurement data can be analysed with different values for parameters like predelay and gate width, without repeating the acquisition. Other useful diagnostic information, such as die-away time and dead time, can also be extracted from this stored data. A second area is the development of “virtual instruments.” These devices, in which the pulse-processing system can be embedded in the neutron counter itself and sends counting data to a PC, can give increased data-acquisition speeds. Either or both of these developments could give rise to the next generation of instrumentation for improved practical neutron-correlation measurements. The paper will describe the rationale for changing to the new technology, give an overview of the hardware and software tools available today and a feedback of the experience gained in the first tests. Associated with the experimental tests, the ESARDA-NDA working group is also performing an intercomparison benchmark exercise on the analysis software for pulse processing.

Calibration of the HB-line active well neutron coincidence counter for measure of LANL 3013 highly enriched uranium product splits

In this paper, the setup, calibration, and testing of the F-Area Analytical Labs active well neutron coincidence counter (HV-221000-NDA-X-1-DK-AWCC-1) in SRNL are described for use in the Savannah River Site (SRS) transuranium metal production facility to enable assay of mixed uranium/plutonium metal product. The instrument was required within a three-month window for availability upon receipt of LANL uranium oxide samples into the SRS facility. Calibration of the instrument in the SRNL nuclear nondestructive assay facility in the range 10–400 g HEU is described. We also report qualification and installation of the instrument for assay of the initial suite of product samples.

Estimation of isotopic composition of plutonium from thermal reactors by neutron coincidence counting through isotopic correlation technique

Correlations have been established between %Eff 240Pu and various plutonium isotopes formed in thermal reactors. Based on these correlations, a method has been developed for the estimation of isotopic composition of plutonium obtained from thermal reactors. The method is simple, fast, non-destructive and finds application for the verification of plutonium isotopic composition in the finished products of known plutonium content. The method could be applied in the nuclear fuel fabrication to verify and confirm the fissile content (239Pu+241Pu) specification. It has also been shown that in principle, similar correlations could be established for Pu obtained from different thermal reactor fuels with reactor specific fitting parameters.

Analysis of A Nuclear Material Accounting Uncertainty for ACP Spent Fuel Standards: MCNPX and ORIGEN-S Code Simulations

The target nuclide to be measured by using a neutron coincidence counter is Cm-244 for the PWR spent fuel since the dominant neutron-emitting source is Cm-244 for the spent fuel. Spent fuel standards (SFS) are usually used to calibrate the neutron coincidence counter for nuclear material measurements of spent fuel. To make the best selection and minimize the measurement uncertainty of neutron coincidence counter for the SFS, expected measurement errors for several kinds of burnups were obtained by using the ORIGEN-S code and the expected singles and doubles count rates were also acquired by using MCNPX code. The expected doubles count errors for 1 pellet of spent fuel are less than 4% for a burnup from 27 to 40 GWD/MtU. Considering everything, a higher burnup of the spent fuel seems to be better to obtain more accurate data from a nuclear material measurement by using a neutron coincidence counter than that of a lower burnup.

Diverse active well neutron coincidence counter utility at the Savannah River National Laboratory

In this paper we describe use of the Aquila active well neutron coincidence counter for nuclear material assays of 235U in multiple analytical techniques at Savannah River Site (SRS), at the Savannah River National Laboratory (SRNL), and at Argonne West National Laboratory (AWNL). The uses include as a portable passive neutron counter for field measurements searching for evidence of 252Cf deposits and storage; as a portable active neutron counter using an external activation source for field measurements searching for trace 235U deposits and holdup; for verification measurements of U-Al reactor fuel elements; for verification measurements of uranium metal; and for verification measurements of process waste of impure uranium in a challenging cement matrix. The wide variety of uses described demonstrate utility of the technique for neutron coincidence verification measurements over the dynamic ranges of 100–5000 g for U metal, 200–1300 g for U-Al, and 8–35 g for process waste. In addition to demonstrating use of the instrument in both the passive and active modes, we also demonstrate its use in both the fast and thermal neutron modes.

Preliminary calibration of the ACP safeguards neutron counter

The Advanced Spent Fuel Conditioning Process (ACP), a kind of pyroprocess, has been developed at the Korea Atomic Energy Research Institute (KAERI). Since there is no IAEA safeguards criteria for this process, KAERI has developed a neutron coincidence counter to make it possible to perform a material control and accounting (MC&A) for its ACP materials for the purpose of a transparency in the peaceful uses of nuclear materials at KAERI. The test results of the ACP Safeguards Neutron Counter (ASNC) show a satisfactory performance for the Doubles count measurement with a low measurement error for its cylindrical sample cavity. The neutron detection efficiency is about 21% with an error of ±1.32% along the axial direction of the cavity. Using two 252Cf neutron sources, we obtained various parameters for the Singles and Doubles rates for the ASNC. The Singles, Doubles, and Triples rates for a 252Cf point source were obtained by using the MCNPX code and the results for the ft8 cap multiplicity tally option with the values of ε, f d, and f t measured with a strong source most closely match the measurement results to within a 1% error. A preliminary calibration curve for the ASNC was generated by using the point model equation relationship between 244Cm and 252Cf and the calibration coefficient for the non-multiplying sample is 2.78×10 5 (Doubles counts/s/g 244Cm). The preliminary calibration curves for the ACP samples were also obtained by using an MCNPX simulation. A neutron multiplication influence on an increase of the Doubles rate for a metal ingot and UO2 powder is clearly observed. These calibration curves will be modified and complemented, when hot calibration samples become available. To verify the validity of this calibration curve, a measurement of spent fuel standards for a known 244Cm mass will be performed in the near future.

Neutron coincidence counter for the verification of PuO 2 cans

A neutron coincidence counter intended for the verification of cans of PuO 2 powder has been designed and optimised using the Monte Carlo radiation transport simulation code–pulse train analyser (MCNP–PTA) which simulates both the neutron transport and the coincidence electronics setup. Severe operational and physical constraints had to be taken into account in designing and optimising the counter as well as in adopting the most adequate strategy for its calibration, validation and application.

A proposed semi-empirical method for 235U mass calibration of the active-well neutron coincidence counter

In this work, a proposed semi-empirical method for 235U mass calibration of the active-well neutron coincidence counter (AWCC) is presented. Samples of low enriched uranium oxide powder in form of cylinders were used to investigate and verify the validity of the method. The calibration process was carried out via two main steps. First, different experimental setup configurations (for the detector and samples) were modeled with the Monte Carlo multi-particle transport code MCNP-4B to calculate, numerically, fission rates in the nuclear material (NM) samples for each configuration. Second, the absolute efficiency of the detector for the real coincidence neutrons was measured using a standard 252Cf point source. The measured efficiency was combined with the calculated fission rates to estimate the real coincidence count rate of the detector due to NM for each setup configuration. A semi-empirical calibration curve relates the 235U mass content in each modeled setup configuration with its corresponding estimated count rate was constructed. The calibration curve was checked against experimental data to verify the proposed method. 235U masses estimated from the calibration curve were found to be in agreement with the measured values within the error limits. An accuracy of better than 7.2% was achieved with an average precision of about 4%.

Development of quantitative analytical methods for the control of actinides in a pyrochemical partitioning process

Summary Advanced nuclear fuel cycles are being developed in order to reduce the long-term radiotoxicity of highly radioactive waste. Pyrochemical partitioning techniques appear particularly attractive for advanced fuel cycles in which the minor actinides are recycled. The electrochemical processes of practical importance are the electrorefining process and the liquid–liquid extraction of transuranic (TRU) elements from fission products using either non-miscible molten metal or molten salt–metal phases. Analytical methods for the accurate assay of actinide elements in these matrices needed to be developed. A quantitative assay is required in order to establish a material balance for process development and – at a later stage – for accountancy and control purposes. To this end radiometric techniques such as energy-dispersive X-ray fluorescence analysis (XRF), neutron coincidence counting (NCC) and high-resolution gamma spectrometry (HRGS) were extensively employed for the quantitative determination of actinides (U, Np, Pu, Am, Cm) in process samples. Comparative analyses were performed using inductively coupled plasma mass spectrometry (ICP-MS). The respective samples were available in small quantities (≅100 mg) either in the form of eutectic salt or in metallic form with Cd, Zr or Bi as major metallic matrix constituents.

Combined calorimetry/neutron coincidence counting/gamma spectrometry (CANEGA) measurements for plutonium mass and isotopic assay

A combined Calorimetry/Neutron Coincidence Counting/Gamma Spectrometry (CANEGA) assay approach offering the potential for improved nondestructive plutonium assay is described. The combination of the six directly measurable quantities as obtained from the three measurement techniques, namely the thermal power P from calorimetry, the effective mass m 240eff of 240Pu from neutron coincidence counting (NCC), and the plutonium isotope ratios 238/239, 240/239 and 241/239 and 241Am/239 from high-resolution gamma spectrometry (HRGS) allow to determine the 242Pu isotope abundance, which otherwise cannot be directly obtained from a HRGS measurement. In this manner the CANEGA measurement approach provides a more complete fingerprint for any plutonium-bearing sample under assay, which in turn helps to improve the overall accuracy of the plutonium mass determination. Pertinent features and performances of the individual and combined measurement techniques are presented and discussed.

A Model for Dead-Time Effects in Coincidence Counting

A physical model describing the production and decay of neutrons in a neutron coincidence counting detector is developed and used to derive the correlation functions for the neutrons in the detector system. The correlation functions serve as the basis for the modeling of dead-time effects for this as well as earlier models. An iterative scheme to correct the dead-time effects from measurements is suggested.

Statistical evaluation of two triggering systems: an application of conditional variance

Previous results related to measurement variances in neutron coincidence counting (NCC) indicate a disagreement between theory and experiment. In NCC, each detected neutron opens two coincidence gates and counts the number of neutrons that arrive within the allotted time period of the first neutron. Similarly, previous results for studying the alarm rate of a broadband radio-frequency collection system (to detect the radio-frequency emitted from nuclear explosions or lightning for example) are shown to lead to poor approximations for some parameter ranges. In a typical example of this type of system, the master trigger could alarm if five or more of eight sub-bands alarm within a specified time window beginning from an alarm from one or more sub-bands. We describe a simple but powerful statistical result involving conditional variance that corrects the previous results in both applications, and we demonstrate the adequacy of the results using real and simulated NCC data and simulated broadband radio-frequency data.

An approach to multiplicity counting for a versatile new sensor for plutonium assay with a very short die-away time, and independent measurements of neutrons and gamma rays

A unique detector design incorporating a 6Li-based capture medium, ZnS scintillator, and wavelength shifting optical fibers is the basis of a new neutron coincidence counter for measurements of plutonium in highly impure residues. The sensor elements have a high efficiency for detecting neutrons and exhibit excellent gamma-ray discrimination based on pulse-shape analysis. The short die-away time of the counter that is based on these detector elements allows coincidence-gate settings shorter than 10μs. This qualifies the technology for measurements of materials with high yields of uncorrelated neutrons from 241Am(α,n) reactions. The characteristics of the new neutron counter will be illustrated with test data from measurements of plutonium, 252Cf, and gamma-ray sources. The integrated electronics design of the new detector also permits the simultaneous but independent measurement of both neutrons and gamma rays. Recent test results that illustrate some unique applications of the sensor's versatility will also be presented.

Study of a new boron loaded plastic scintillator (revised)

Neutron detection research using plastic scintillators has recently led to instrument development for several applications and particularly for nuclear material characterization. In particular, plastic scintillators exhibit a fast time response and a high efficiency for fast neutrons spectrometry. The use of boron loaded plastic scintillators has been also proposed in order to reduce dead time for fission neutron coincidence counting. We report here on the study of a new and low-cost boron loaded plastic scintillator. Further, a simulation study was performed, including light production and transport in this kind of scintillator. Its characteristics have been measured in terms of light response, photon and neutron induced signals. An experiment has been carried out using the ion beam of a Van de Graaff accelerator at Saclay so as to determine the Birks parameters of the scintillators. These parameters were obtained for proton (between 1-2 MeV) and alpha particles (between 3.1 MeV and 5.4 MeV). The correlation between the experimental data and the simulation predictions has been studied for the neutron response of these scintillators. In addition, performances of this scintillator are compared with those of commercially available boron loaded plastic scintillator (BC454 from Bicron, Newbury, OH).

The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as ∼10 5 Rads/h. Using limited facility information and multiple measurements along the length of spent-fuel assemblies, the combined measurement and facility declaration error is ∼8%. A simplified one-point measurement procedure leads to a combined measurement and facility declaration error of ∼13%.

Study of a new boron loaded plastic scintillator

Neutron detection research using plastic scintillators has recently led to instrument development for several applications and particularly for nuclear material characterization. In particular, plastic scintillators exhibit a fast time response and a high efficiency for fast neutrons spectrometry. The use of boron loaded plastic scintillators has been also proposed in order to reduce dead time for fission neutron coincidence counting. We report on the study of a new and low-cost boron loaded plastic scintillator. Further, a simulation study was performed, including light production and transport in this kind of scintillator. Its characteristics have been measured in terms of light response, photon and neutron induced signals. An experiment has been carried out using the ion beam of a Van de Graaff accelerator at Saclay so as to determine the Birks parameters of the scintillators. These parameters were obtained for proton (between 1-2 MeV) and alpha particles (between 3.1 MeV and 5.4 MeV). The correlation between the experimental data and the simulation predictions has been studied for the neutron response of these scintillators. In addition, the performance of this scintillator are compared with those of a commercially available boron loaded plastic scintillator (BC454 from Bicron, Newbury, OH).

MCNP–REN: a Monte Carlo tool for neutron detector design

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP–REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP–REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel were taken with the Underwater Coincidence Counter, and measurements of highly enriched uranium reactor fuel were taken with the active neutron interrogation Research Reactor Fuel Counter and compared to calculation. Simulations completed for other detector design applications are described. The method used in MCNP–REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.

Evaluation of a fast neutron coincidence counter for the measurements of uranium samples

A fast neutron coincidence counter using BC454/BGO phoswich detectors has been evaluated for the purpose of rapid verification measurements of uranium items. This counter uses custom electronics to identify and count coincidence neutrons in the presence of background radiation. Measurements of uranium standards were performed to evaluate the counter. This counter is successful in measuring uranium items but has a low efficiency that results in minimal improvement over current technology. An optimized counter can be built with better performance capabilities, but it is recommended that newer technologies be used instead.

The measurement of passive neutron multiplicity counter gate utilisation factors and comparisons with theory

In a recent paper, a new method of calculating the doublets and triplets gate utilisation factors f 2 and f 3 for passive neutron multiplicity counters has been developed resulting in the MCF computer code. In this paper we arrange the basic multiplicity equations and generate nuclear data parameter combinations to enable f 2, f 3, f 2/ f 3 and f 2 2/ f 3 to be derived from measurements taken using a calibrated 252Cf source and multiplicity shift-register electronics. The two approaches have been compared with reference to Harwell Instruments Ltd. N95 neutron coincidence counting measurement chamber operated both with and without an internal cadmium liner. The Monte Carlo code MCNP™ was used to calculate neutron capture time distributions and these were fed into the MCF code to produce theoretical gate utilisation factors to compare with the experimental data. Pre-delays of 2 and 4 μs and coincidence gate widths between 16 and 240 μs were investigated. The predicted values agreed with those measured within 1.5% and 5% for the f 2 and f 3 results, respectively. These results confirm the power of the new theoretical treatment in accurately predicting gate utilisation factors in an absolute sense and with confidence at the design stage. Relative predictions can be made with greater accuracy.