Neutron Coincidence Counting Research Articles

Probabilistic radwaste characterization: Findings of a multi-method multi-mockup exercise using interpolation-based surrogate efficiencies

In this paper, we present the lessons learned from a multi-method, multi-mockup probabilistic radiological characterization exercise conducted within the EU project CHANCE “Characterization of Conditioned Nuclear Waste for its Safe Disposal in Europe”. A Bayesian approach that accounts for uncertainty in the measurement efficiencies is used to interpret combinations of (i) open geometry gamma spectrometry (OGGS), (ii) passive neutron coincidence counting (PNCC) and (iii) large-volume calorimetry (LVC) measurements, performed on five different combinations of mockup drum and source position. The used approach treats uncertainty in the measurement efficiencies by doing multilinear interpolation between reference efficiencies representing potential “end-member” matrices with respect to both composition and density, for the considered drum. The end-member proportions are then jointly inferred with the other unknowns. Our results indicate that the success of this method critically depends on the accuracy of the set of reference end-member efficiencies obtained from physics-based Monte Carlo particle transport modeling. We find that Bayesian inversion of OGGS data alone can recover reasonably accurate nuclide masses, with relatively high precision and low bias. Furthermore, when little prior information about the drum is available then jointly inverting OGGS and PNCC data is found to allow for a much higher precision compared to inverting OGGS data alone. In addition, calorimetry appears to be a promising method and brings some added value to OGGS as well in certain cases. In some setups and for some nuclides, errors in the reference efficiencies cause the marginal posterior distribution to be fairly biased. We thus design a simple approach to increase posterior uncertainty (that is, decrease precision) such that the true mass values are included in the bulk of the posterior distribution, in a controlled way. The approach consists of multiplying the interpolated efficiencies with a series of multipliers, that are jointly inferred under a relatively strong prior.

Fast neutron coincidence counter using scintillators for safeguards verification of the TRU/RE waste

Fast neutron coincidence counter using scintillators for safeguards verification of the TRU/RE waste

Feasibility of neutron coincidence counting for spent TRISO fuel

High-temperature gas reactors rely on TRIstructural-ISOtropic (TRISO) fuel for enhanced fission product retention. Accurate fuel characterization would improve monitoring of efficient fuel usage and accountability. We developed a new neutron multiplicity counter (NMC) based on boron coated straw (BCS) detectors and used it in coincidence mode for 235U assay in TRISO fuel. In this work, we demonstrate that a high-efficiency version of the NMC encompassing 396 straws is able to estimate the 235U in used TRISO-fueled pebbles or compacts with a relative uncertainty below 2.5% in 100s. We performed neutronics and fuel depletion calculation of the HTR-10 pebble bed reactor to estimate the neutron and gamma-ray source strengths of used TRISO-fueled pebbles with burnup between 9 and 90GWd/t. Then, we measured a gamma-ray intrinsic efficiency of 10−12 at an exposure rate of 340.87R/h. The low gamma-ray sensitivity and high neutron detection efficiency enable the inspection of used fuel.

Neutron coincidence counters for disarmament verification: Experiments and modeling

Neutron coincidence counters for disarmament verification: Experiments and modeling

Dead time correction of Feynman sampled rates suitable for neutron coincidence counting

Neutron coincidence counting (NCC) based on shift-register logic (SRL) is extensively used by international nuclear safeguards inspectorates to verify special nuclear material inventories, particularly Pu. An alternative autocorrelation pulse counting analysis technique is founded on Feynman sampling and makes use of the well-known Feynman-Y statistic. Although the latter approach has been widely adopted by the reactor noise and sub-criticality communities it has not been used by the nonproliferation community. Both techniques may be interpreted using the same theoretical model and so should extract the same fundamental information about the detected neutron pulse train, namely the first and second reduced factorial multiplets that are determined by the joint properties of the measurement item and the surrounding external neutron detector. In the nuclear safeguards domain, a semi-empirical dead time treatment is conventionally applied to the trigger (singles) and doubles rates and, as already noted, the Feynman-Y method is not typically applied. In this article, we revisit the dead time treatment of Robba et al. (1983) for the Feynman sampling method. We show so that a dead time corrected multiplet analysis, familiar to safeguards practitioners, can be extracted. This is a first and necessary step to enabling a comparison of the relative performance of these two autocorrelation techniques on a like basis for deeply subcritical measurement items, which is the domain of concern to international nuclear safeguards inspectorates. From the analytical derivations it emerges that the Robba formalism, with minor corrections, and the commonly used NCC shift-register logic dead time correction formalism yield numerically equivalent corrections for practical purposes over the operational range of safeguards interest.

Design of MICADO advanced passive and active neutron measurement system for radioactive waste drums

In the frame of the MICADO H2020 project, a passive and active neutron measurement system is being developed to estimate the nuclear material mass inside legacy waste drums of low and intermediate radioactivity levels. Monte-Carlo simulations have been performed to design a new modular and transportable neutron system, with the main objective to reach a good tradeoff between the performances in passive mode, i.e. neutron coincidence counting, and in active interrogation mode with the Differential Die-away Technique. Different designs are compared, which mainly differ in their moderation materials, graphite and polyethylene. This parametric study allowed us to define a prototype taking into account practical constraints in view of its final implementation in a wide range of in-situ locations and nuclear facilities. The total neutron detection efficiency of the prototype is 6.75%, as calculated for an empty drum, i.e. without waste matrix. The detection limit in terms of nuclear material equivalent mass have also been estimated based on assumptions for a homogeneous distribution of nuclear materials inside the drum, filled with four types of matrices covering the range of nuclear waste drums defined in the frame of the project. The most favorable matrix is made of stainless steel in passive mode and of polyethylene in active mode, with an apparent density of 0.7 g cm−3 and 0.1 g cm−3, respectively. The calculated mass detection limits are respectively 68 mg of 240Pu, 62 mg of 235U and 39 mg of 239Pu. The most penalizing matrix is made of polyethylene with an apparent density of 0.7 g cm−3, which leads to a mass detection limit of 519 mg of 240Pu in passive mode, and 564 mg of 235U or 349 mg of 239Pu in active mode. Measurement time is 30 min for both passive and active modes. Next steps will be a complete investigation of matrix effects based on intensive Monte-Carlo calculations and an experimental design to figure out the appropriate corrections. Experiments will also be conducted at CEA Cadarache Nuclear Measurement Laboratory with the construction and the assembly of the neutron system prototype, and the measurement of mock-up drums filled with different matrices.

Neutron Coincidence Measurements and Monte Carlo Modelling of Waste Drums Containing Reference Nuclear Material

Within the EC-funded CHANCE project several non-destructive techniques are being considered for the assay of waste bearing drums. Such techniques include calorimetry, gamma-ray spectrometry and neutron coincidence counting. The aim is to quantify uncertainties on the inventory of radionuclides, and how these are potentially reduced by combining the signatures from different techniques in the data analysis. In this framework, neutron coincidence measurements were carried out with two slab counters based on 3He detectors coupled to shift register electronics. Such a system consists of two identical slabs with 6 detectors each, and is transportable, rather compact and flexible in terms of sizes and geometries that can be measured. With this system three 200 L drums containing certified reference nuclear material and different filling materials were measured. The certified nuclear material was in the form of 21 pellets of mixed oxide of U and Pu with a total mass of about 10.5 g; in addition, a single pellet of about 10.05 g was also available. The pellets could be placed in predefined positions within the drum in a reproducible way. The geometry and composition of the three drums was well characterized and consisted of Ethafoam, a mixture of Ethafoam, stainless steel and PVC, and mortar with an inner core of extruded polystyrene. The measurement setup was arranged such that the drum was placed between the two slab counters. The positions of the slab counters relative to the drum were accurately measured before each measurement, and a dedicated system was used to minimize the uncertainty on the detector positioning. The measurement data were first analysed by applying the point model of Hage and the mass of nuclear material in the drum was determined from the rate of totals and reals and the radionuclide composition. Due to the fact that not all the point model conditions were met, we found that the point model overestimates the mass up to about 50%. In addition, a Monte Carlo model of the measurement geometry was developed using the MCNP code. The model was used to determine a calibration factor between the reals rate and the mass of the sample. Measurements with a calibrated 252Cf source were used to verify the model. With a Monte Carlo based approach the mass of the mixed oxide pellets is within a few percent from the nominal values, except for strongly asymmetrical configurations where the deviation is up to about 20%. The results reveal the importance of an accurate background correction and of accounting for surrounding materials of the building such as walls, floor and ceiling in the Monte Carlo model.

Development and evaluation of a list mode neutron coincidence collar for spatial response measurements of fresh fuel assemblies

The capabilities of current safeguards neutron instrumentation can be expanded by acquiring and analyzing data in new ways. A new List Mode Collar (LMCL) was created by retrofitting a traditional safeguards neutron coincidence counting system, the Neutron Coincidence Collar, with newly developed preamplifiers on each of its 18 3He tubes in active interrogation mode. By applying list mode data acquisition (LMDA) and analysis, a signal from each 3He tube can now be recorded and analyzed to determine the neutron count rate measured in each individual 3He tube. Further, the count rate distribution across all 3He tube locations in the system can be determined. This is the first time that a full list mode method has been implemented using a traditional neutron coincidence collar and spatial response data has been obtained.The objective of this research was to develop and evaluate a laboratory demonstration list mode neutron coincidence collar (LMCL) that can extract a greater number of useful signatures than are currently generated using the current collar. A new analysis method called the “List Mode Response Matrix” has also been developed for use in combination with these measurements to improve the detection of missing fuel pins (partial defects) during the nondestructive assay of fresh nuclear fuel assemblies, although analysis algorithms will be reported separately from this experimental evaluation. In this paper, 18 channel spatial response data are presented for a 252Cf source moving within the LMCL detector cavity. LMDA, combined with the addition of multiple preamplifiers, facilitates this capability by increasing the number of signals that can be measured simultaneously; allowing for in-depth analysis of neutron coincidence events to determine the location and distribution of nuclear material within the measurement geometry. This approach enables a spatial response measurement to form a “pattern” or item signature unique to the fissioning source’s distribution; in the case of the LMCL, the distribution of fresh low-enriched uranium fuel within the assembly. A major benefit provided by LMDA is that this can all be achieved from a single measurement pulse train in offline analysis.To experimentally demonstrate this concept, a Mirion Technologies (Canberra) model JCC-71 Neutron Coincidence Collar has been retrofitted with new electronics designed at Oak Ridge National Laboratory. This paper presents the results from the experimental evaluation of the laboratory demonstration system. The results indicate equivalent performance of the LMCL to the original collar in addition to improved spatial response capabilities, while maintaining the original system footprint.

A Direct Nondestructive Assay for the Pu of U/TRU Ingot in Pyroprocessing Using 244Cm Neutron Albedo Reactivity Technique and Its Error

The transuranic (TRU) ingot is considered to be the most prominent target material of pyroprocessing in terms of safeguards since it contains almost all of the Pu of the feed spent fuel. Due to the high density, excessively high neutron emission rates, and high neutron multiplication of the U/TRU ingot, it is impractical to apply gamma-ray spectroscopy or neutron coincidence counting techniques to the quantification of the Pu content of the U/TRU ingot. Since the passive neutron albedo reactivity (PNAR) technique is known to be sensitive to the total fissile mass of target material and the uncertainty of its singles Cd ratio is independent of the accidental coincidence coming from the high neutron emission rate, the capability of the PNAR technique for the quantification of the Pu content of the U/TRU ingot has been investigated using the MCNPX code with a spent fuel library with 81 different cases of various kinds of initial enrichment, burnup, and cooling time. The MCNPX simulation results for the Cd ratio versus Pu content of the U/TRU ingot show the maximum error in the Pu mass between the linear fit and the real Pu content in the U/TRU ingot is 2.14% for 4.5 wt% initial enrichment cases. The results of this study show that the PNAR technique can be one possible method for the direct nondestructive assay for the Pu of the U/TRU ingot.

Performance Assessment of Amplification and Discrimination Electronic Devices for Passive Neutron Measurements

The assessment of fissile material mass is a key requirement to enhance radioactive waste management and to ensure a high level of safety in the nuclear industry. The assessment of plutonium fissile mass by passive coincidence measurements is usually obtained by detecting the neutrons generated by the spontaneous fissions of Pu isotopes. The detection and the quantification of a fissile mass in the radioactive waste can also be carried out using active neutron interrogation with a pulsed D-T neutron generator. The 14-MeV neutrons are moderated to induce fissions in the uranium and plutonium-contaminated waste. The emitted neutron signal can be analyzed by considering their detection time after the generator pulse. Data are measured according to the principles of the neutron measurement techniques. As proportional counters filled with 3He gas display high neutron detection efficiency and good gamma-ray immunity, they are the reference detector for passive neutron coincidence counting. A charge preamplifier or a current amplifier, depending on applications, collects the electric pulse produced by neutron interaction in the 3He gas, and a threshold discriminator produces a logic pulse used for neutron counting. This article describes the performance assessment of different commercially-available electronics from Mirion Technologies, Precision Data Technology (PDT), Mesytec, as well as MONACO electronics originally developed by CEA LIST for fission chamber measurements in experimental reactors. Comparative passive neutron measurements are carried out with these electronics at CEA/DEN Nuclear Measurement Laboratory in Cadarache. Overall, PDT and Mesytec electronics show similar detection efficiency as the ACH-NA98 charge amplifier, which is commonly used in our laboratory for such applications. However, MONACO electronics have a lower detection efficiency, similar to the Mirion 7820 current amplifier used in specific high-count rate applications. An optimization of MONACO settings would probably be necessary to adapt to 3He counters instead of fission chambers.

High-Resolution Gamma Spectrometry of a Plutonium Bearing Waste Drum With High-Energy Reaction-Induced Gamma Rays

In the framework of the radioactive waste drum characterization using neutron coincidence counting, the Nuclear Measurement Laboratory of French Alternative Energies and Atomic Energy Commission (CEA) Cadarache is studying plastic scintillators as an alternative to ideal but costly 3He gas proportional counters. Plastic scintillators are at least five times cheaper for the same detection efficiency, and in addition, they detect fast neutrons three orders of magnitude more quickly than 3He detectors. However, they are sensitive to gamma rays with no pulse shape discrimination abilities for large detection volumes, which implies the necessity to identify precisely gamma background sources that may affect the useful signal. This article presents a detailed analysis of the gamma-ray spectrum of a radioactive waste drum containing glove box filters contaminated by plutonium dioxide. Classical gamma emissions following alpha and beta disintegrations are identified, and also those accompanying inelastic scattering (n, $\text{n}^{\prime }$ ) or radiative capture (n, $\gamma$ ) reactions in the whole waste drum, and ( $\alpha $ , n) or ( $\alpha $ , p) reactions in the filtration media, which can lead to neutron-gamma coincidences parasitizing useful coincidences from plutonium spontaneous fissions.

Field testing of a fast-neutron coincidence collar for fresh uranium fuel verification

The Fast-Neutron Coincidence Collar (FNCL) is an active non-destructive assay system for the measurement of uranium-235 in fresh nuclear fuel assemblies using the neutron coincidence counting (NCC) approach. Unlike traditional NCC instruments, such as the Uranium Neutron Collar (UNCL), the FNCL contains liquid scintillator to detect fast neutrons (>0.1 MeV) with nanosecond-level time resolution. The subsequent reduction of the NCC coincidence window by a factor of one thousand (from microseconds to nanoseconds) allows the uncertainties associated with accidental neutron coincidences to be virtually eliminated resulting in precise and accurate assays of fresh fuel assemblies (FFAs) in significantly less time than alternatives. The measurement of fast neutrons also minimizes the systematic uncertainties due to burnable neutron poisons commonly present in the fuel matrix.The FNCL was tested at two nuclear fuel fabrication facilities in Brazil and the Republic of Korea with continued support from the two States. 39 measurements of FFAs were obtained with varying geometry, enrichment and burnable neutron poison content. These field test measurements and conclusions drawn pertaining to the FNCL performance are discussed.

Estimation of fissile material content in irradiated In-Pile Sections using neutron coincidence counters

A set of In-Pile Sections (IPS) has been irradiated in the BR2 reactor at SCK•CEN in Belgium during the 1970’s and 1980’s. The primary goal of the IPS was to replicate the thermo-hydraulic loop of a sodium-cooled fast reactor in order to study severe accident scenarios. The top part of the IPS contained the sodium-cooled loop whereas the lower part contained the fuel element. Due to the experimental conditions, the rupture of the fuel pins contained in the IPS occurred and fuel fragments may have been deposited in the rest of the IPS loop. The part of the IPS containing the fuel pins has been cut from the rest of the IPS and underwent post-irradiation examinations at specialized EU laboratories, while the top parts remained stored at SCK•CEN. To prepare for future transport, dismantling and conditioning, a reliable estimation of the total fissile content in the stored parts of the IPS is indispensable. In this framework, two IPS were measured with a Canberra WM3400 neutron coincidence counter with customized electronics. The measurements of the IPS were challenging due their length (roughly 6 m) and intense gamma-ray radiation background. For each IPS an axial scan was carried out with a series of short measurements (600-700 s each) recording the Totals rate and Reals rate. Based on the results of the axial scans, measurements with longer measurement time were conducted for the axial positions with the larger values of Reals rates. A system of equations was then established to quantify the 240Pu content in the different sections of the IPS from the Reals rates in each measurement position and account for cross-talk between the neutron emission associated to the different sections. A set of Monte Carlo simulations was carried out to estimate the probability to record a Real count in the detector due to spontaneous fission events occurring in a given section of the IPS. The 240Pu content in each section of the IPS was calculated by combining the measured Reals rates and the detection probabilities calculated with the simulations. The total fissile content in the IPS was then determined with scaling factors based on burnup calculations for the irradiated fuel assemblies in the IPS. The results indicate that both IPS measured with the neutron coincidence counters have a fissile content lower than the limit for transport. It is expected that the envisaged segmentation of the IPS in shorter sections required to fit into 200L drums will provide an additional safety margin on this limit.

Improved electronics for 3He based neutron counters

Several non-destructive assays techniques have been developed for the measurement of fissile materials in the fields of dismantling, decommissioning, nuclear security, and nuclear safeguards. Among these techniques, neutron coincidence counting is based on the detection of time-correlated neutrons from induced and spontaneous fissions. 3He Tubes have been the primary choice for neutron coincidence counting due to their high detection efficiency, rather low sensitivity to gamma-rays and proven field reliability. This paper covers the implementation of a new electronic setup to a Canberra WM3400 neutron coincidence counter. First we describe the properties of the used detectors, with focus on the characteristics of the default electronics and highlight its limitations such as the high input capacitance, short shaping time and the necessity for selected tubes. We then propose the new electronic setup to overcome these limitations. This setup includes a dedicated preamp for every tube , the possibility to adjust for gain differences between the tubes and a better optimised shaping time for 3He detectors. We carried out measurements with the two electronic systems to compare their performances in terms of gamma-ray sensitivity, efficiency and die-away time. The gamma ray sensitivity was measured with calibrated 137Cs and a 60Co sources at the Laboratory for Nuclear Calibration of the Belgian Nuclear Research Centre with dose rates between 10 μSv/h and 50 mSv/h. Measurements with a 252Cf source were used to determine the die-away time of the system and the total measurement efficiency for the considered geometry. The measurements showed that, with the default electronics, neutron count-rates are already affected by gamma radiation at a dose rate of 10÷30 μSv/h. On the other hand the neutron coincidence counter equipped with the new electronics proved to be insensitive to gamma-radiation up to a dose rate of at least 20 mSv/h. The high-voltage set with the new electronics is lower than in the case of the default electronics and is within the range recommended by the tubes manufacturer. The die-away time was not affected by the used electronics. A reduction of about 20% in the neutron detection efficiency due to the used discriminator threshold was observed.

Study of gamma-ray background noise for radioactive waste drum characterization with plastic scintillators

In the framework of the radioactive waste drum characterization using neutron coincidence counting, the Nuclear Measurement Laboratory of CEA Cadarache is studying plastic scintillators as an alternative to ideal but costly 3He gas proportional counters. Plastic scintillators are at least 5 times cheaper for the same detection efficiency, and in addition, they detect fast neutrons about three orders of magnitude faster than 3He detectors. However, they are sensitive to gamma rays, which implies the necessity to identify precisely gamma background sources that may affect the useful signal. This paper presents a detailed analysis of the gamma-ray spectrum of a radioactive waste drum containing glove box filters contaminated by plutonium dioxide. Gamma emissions accompanying inelastic scattering (n,n’) and (α,n) reactions that can lead to neutron-gamma coincidences parasitizing useful coincidences from plutonium spontaneous fissions are identified. Some of these parasitic gamma rays having energies up to several MeV, we plan to reject high-energy scintillator pulses with an electronics rejection threshold above 1 MeV, which should preserve the major part of useful fission neutron pulses.

Performance assessment of amplification and discrimination electronic devices for passive neutron measurements

The knowledge of the fissile material mass is a key challenge to enhance radioactive waste management and to ensure a high level of safety in nuclear industry. Data is analyzed according to the principles of the neutron measurement techniques. As proportional counters filled with 3He gas display high neutron detection efficiency and a good gamma-ray discrimination, they are the reference detector for passive neutron coincidence counting. A charge preamplifier or a current amplifier, depending on applications, collects the electric pulse produced by neutron interaction in the 3He gas and a threshold discriminator produces a logic pulse used for neutron counting. This paper describes the performance assessment of different commercially available electronics from Mirion Technologies, Precision Data Technology (PDT), Mesytec, as well as MONACO electronics originally developed by CEA LIST for fission chamber measurements in experimental reactors. Comparative passive neutron measurements are carried out with these electronics at CEA/DEN Nuclear Measurement Laboratory in Cadarache. Overall, PDT and Mesytec electronics show similar detection efficiency as the ACH-NA98 charge amplifier, which is commonly used in our laboratory for such applications. However, MONACO electronics have a lower detection efficiency, similar to Mirion 7820 current amplifier used in specific high-count rate applications. An optimisation of MONACO settings would probably be necessary to adapt to 3He counters instead of fission chambers.

Results of the Swedish spent fuel measurement field trials with the Differential Die-Away Self-Interrogation Instrument

Differential Die-Away Self-Interrogation (DDSI) is a method by which the characteristic die-away of neutrons from spontaneous and induced fissions is used to characterize a spent nuclear fuel assembly. A nondestructive assay (NDA) instrument was built at Los Alamos National Laboratory to implement and test the DDSI method. The DDSI instrument contains 3He detectors which measure thermal neutrons, and the time and location of detection of each neutron is recorded via list-mode data acquisition. The instrument was sent to the Clab interim storage facility in Sweden for measurement and characterization of 50 spent pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. The result was over 40 h of neutron list-mode data from a wide variety of fuel assemblies with high enough efficiency to perform neutron coincidence counting, i.e. detection of time-correlated neutrons from fission. Analysis algorithms for characterization of the fuel assemblies were tested on the Swedish spent fuel dataset. Using the measured data, multiplication, fissile mass, initial enrichment, burnup, and total plutonium mass were determined in the 50 assemblies with root mean square errors ranging from 1.5% for PWR multiplication to 11.4% for BWR fissile mass. The results in this work demonstrate that the DDSI concept is capable of characterizing spent power reactor fuel with levels of accuracy that are compatible with the requirements and objectives of various applications such as safeguards verification or facility material control and accounting.

Methods for diagnosing and quantifying double pulsing in a Uranium Neutron Collar system using shift register logic

Double pulsing in neutron coincidence counters is the result of a pulse processing chain- 3 He proportional counter timing incompatibility known to nondestructive assay system designers. It is not currently widely acknowledged, or accounted for, in the user community. However, it is gaining attention as list mode data acquisition and analysis becomes more commonly used for system diagnostics, revealing features that have been overlooked by historic timing gate selection using traditional shift register data acquisition methods. Double pulsing increases the apparent number of measured neutron events from a source. Therefore, it may contribute to a falsely increased count rate if present at the operational high voltage used when assaying samples, even with set predelays. The authors have previously studied the effects of double pulsing, using list mode data acquisition and analysis on neutron pulse trains in three common neutron coincidence counting systems used in routine measurements for international safeguards: a Canberra Industries JCC-71 Neutron Coincidence Collar, a variant on the Canberra Industries JCC-51 Active Well Neutron Coincidence Counter (the Large Volume Active Well Neutron Coincidence Counter), and an AnTech Inc. N2071 Neutron Coincidence Collar. This non-ideal behavior was isolated to the Amptek A111 Charge Sensitive Preamplifier & Discriminator chip used in both the Canberra Industries and Antech systems. The double pulsing fraction was calculated in post-analysis for various high voltage settings in these A111-based systems, assuming no predelay settings. In this work, the authors expand upon this identification and analysis to make it more translatable between list mode data acquisition and analysis and shift register-based analysis. By investigating the double pulsing fractions in a JCC-71 Uranium Neutron Collar as a function of predelay settings, as well as describing and performing alternative tests and analyses to diagnose and quantify the double pulsing using shift register logic, this work hopes to complete the picture of double pulsing identification and analysis.

High order fast neutron multiplicity measurement equations based on liquid scintillation detector

Fast neutron multiplicity counting (FNMC) analysis method is a new nondestructive analysis method for nuclear materials. It uses a fast neutron detector array to detect the neutrons emitted by the sample. The quality of plutonium in the sample is obtained by recording the neutron coincidence counting of fast neutron multiplicity. At present, the first three multiplets widely studied including singlets, doublets and triplets can no longer meet the needs of the research. In this paper, The derivation process of fast neutron multiplicity measurement equation for liquid scintillation detector is studied in detail based on the basic principle, neutron counting method, by using the methods of the factorial moments, probability generating functions and parameter estimation method, and considering the influence of scattering crosstalk. Finally, the fast neutron multiplicity measurement equation including singlets, doublets, triplets, quadruplets and pentuplets are established according to the parameter estimation method and the two kinds of solution methods of this equation are given. The work in this paper lays a theoretical and analytical foundation for the research of fast neutron multiplicity measurement system based on liquid scintillation detector.

Measurement of Tensor Analyzing Power Components of the γd→pnπ0 Reaction

The results of measurements of three tensor analyzing power components of the reaction of incoherent photoproduction of a π0-meson on a deuteron in the ranges of proton and neutron energies, 15–200 and 15–150 MeV, respectively, are presented. The experiment is performed on an internal tensor-polarized gaseous deuterium target of a VEPP-3 electron storage ring using the method of proton and neutron coincidence counting. The results of measurements of tensor analyzing power components are compared with those obtained by statistical modeling performed within a theoretical model.