Nondestructive Assay Techniques Research Articles

Influence of the neutron transport tube on neutron resonance densitometry

Neutron Resonance Densitometry (NRD) is a non-destructive assay technique of nuclear materials in particle-like debris that contains various materials. An aim of NRD is to quantify nuclear materials in a melting fuel of Fukusima Daiichi plant, spent nuclear fuel and annihilation disposal fuel etc. NRD consists of two techniques of Neutron Resonance Transmission Analysis (NRTA) and Neutron Resonance Capture Analysis (NRCA) or Prompt Gamma-ray Analysis (PGA). A density of nuclear material isotopes is decided with NRTA. The materials absorbing a neutron in a wide energy range such as boron in a sample are identified by NRCA/PGA. The information of NRCA/PGA is used in NRTA analysis to quantify nuclear material isotopes. A neutron time of flight (TOF) method is used in NRD measurements. A facility, consisting of a neutron source, a neutron flight path, and a detector is required. A short flight path and a strong neutron source are needed to downsize such a facility and put NRD into practical use. A neutron transport tube covers a flight path to prevent noises. In order to investigate the effect of neutron transport tube and pulse width of a neutron source, we carried out NRTA experiments with a 2-m short neutron transport tube constructed at Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC), and impacts of shield of neutron transport tube and influence of pulse width of a neutron source were examined. A shield of the neutron transport tube reduced a background and had a good influence on the measurement. The resonance dips of 183 W at 27 eV was successfully observed with a pulse width of a neutron source less than 2 μs.

Performance evaluation of PRIDE UNDA system with pyroprocessing feed material

The PRIDE (PyRoprocessing Integrated inactive DEmonstration) is an engineering-scale pyroprocessing test-bed facility that utilizes depleted uranium (DU) instead of spent fuel as a process material. As part of the ongoing effort to enhance pyroprocessing safeguardability, UNDA (Unified Non-Destructive Assay), a system integrating three different non-destructive assay techniques, namely, neutron, gamma-ray, and mass measurement, for nuclear material accountancy (NMA) was developed. In the present study, UNDA's NMA capability was evaluated by measurement of the weight, 238U mass, and U enrichment of oxide-reduction-process feed material (i.e., porous pellets). In the 238U mass determination, the total neutron counts for porous pellets of six different weights were measured. The U enrichment of the porous pellets, meanwhile, was determined according to the gamma spectrums acquired using UNDA's NaI-based enrichment measurement system. The results demonstrated that the UNDA system, after appropriate corrections, could be used in PRIDE NMA applications with reasonable uncertainty. It is expected that in the near future, the UNDA system will be tested with next-step materials such as the products of the oxide-reduction and electro-refining processes.

Sensitivity analysis of high resolution gamma-ray detection for safeguards monitoring at natural uranium conversion facilities

Under the policies proposed by recent International Atomic Energy Agency (IAEA) circulars and policy papers, implementation of safeguards exists when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed in previous work to develop and validate gamma-ray nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP). In particular, uranyl nitrate (UO2(NO3)2) solution exiting solvent extraction was identified as a key measurement point (KMP). Passive nondestructive assay techniques using high resolution gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely manner. Building upon the aforementioned previous validation work on detector sensitivity to varying concentrations of uranyl nitrate via a series of dilution measurements, this work investigates detector response parameter sensitivities to gamma-ray signatures of uranyl nitrate. The full energy peak efficiency of a detection system is dependent upon the sample, geometry, absorption, and intrinsic efficiency parameters. Perturbation of these parameters translates into corresponding variations of the 185.7keV peak area of the 235U in uranyl nitrate. Such perturbations in the assayed signature impact the quality or versatility of the safeguards conclusions drawn. Given the potentially high throughput of uranyl nitrate in NUCPs, the ability to assay 1 SQ of material requires uncertainty «1%. Accounting for material self-shielding properties, pipe thickness, and source-detector orientation is instrumental in determining the robustness of gamma-ray detection in the process monitoring of uranyl nitrate in NUCPs. Monte Carlo models and ray-tracing models were employed to determine the sensitivity of the detected 185.7keV photon to self-shielding properties, pipe thickness, and source-detector geometry. Considering the implementation of the detection of 1 SQ, diversion of 1 SQ becomes essentially undetectable given the systematic uncertainty, in addition to considerations such as propagating uncertainties due to pipe offset/position, as well as minor variations in pipe thickness. Consequently, pipe thickness was the most sensitive variable in affecting full energy efficiency of the 185.7keV signature peak with up to 8% variation in efficiency for ±0.5mm changes in Schedule 40 304L stainless steel piping. Furthermore, computation of the attenuation correction factor of the uranyl nitrate solution [CF(AT) (i.e. εsample)] using Parker's method using with the approximation for the geometrical factor κ≈π/4 was validated through experimental, Monte Carlo and ray-tracing calculations for a uranyl nitrate filled transfer pipe segment. Quantifying sensitivity in detector position, as well as voiding effects due to bubbly flow or laminar flow with an air gap in the uranyl nitrate becomes increasingly important as considerations from (static) design-scale measurements translate into (dynamic) field operations tests.

Principle validation of nuclear fuel material isotopic composition measurement method based on photofission reactions

ABSTRACTWe have developed a new nondestructive assay technique based on the photonuclear reaction, which is aimed at measuring the isotopic composition of nuclear fuel materials without relying on their self-generated neutron information. This methodology enables measurement of the number of neutrons produced by the photofission reactions at different specific photon energies, and only information on the relative counts and not the absolute values or energies, which are considerably affected by backgrounds, are required for the enrichment induction according to a mathematical process. The present methodology estimated the 235U enrichment value with 8% accuracy when the cross section uncertainty is 5%.

Calibration Method of the Tomographic Gamma Scan Techniques Available for Accurately Characterizing 137Cs from 110mAg Interference

The Tomographic Gamma Scan (TGS) technique partitions radioactive waste drums into voxels and assays both the density and concentration of radioactivity for each voxel thus providing for improved accuracy, when compared to the traditional Non-Destructive Assay(NDA) techniques. It could decrease the degree of precision measurement since there is a trade-off between spatial resolution and precision. This latter drawback is compensated by expanding the Region of Interest (ROI) that differentiates the full energy peaks, which, in turn, results in an optimized degree of precision. The enlarged ROI, however, increases the probability of interference among those nuclides that emit energies in the adjacent spectrum. This study has identified the cause of such interference for the reference nuclide of the TGS technique, (661.66 keV, half-life 30.5 years), to be (657.75 keV, half-life 249.76 days). A new calibration method of determining the optimized ROI was developed, and its effectiveness in accurately characterizing and eliminating the interference was further ascertained.

Validation of gamma-ray detection techniques for safeguards monitoring at natural uranium conversion facilities

Recent IAEA circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under the revised policy, IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed to develop and validate concepts of nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP). In particular, uranyl nitrate (UO2(NO3)2) solution exiting solvent extraction was identified as a key measurement point (KMP), where gamma-ray spectroscopy was selected as the process monitoring tool. The Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility at Oak Ridge National Laboratory was employed to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in an NUCP. Nondestructive assay techniques using gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely way. This work investigated gamma-ray signatures of uranyl nitrate circulating in the UNCLE facility and evaluated various gamma-ray detector sensitivities to uranyl nitrate. These detector validation activities include assessing detector responses to the uranyl nitrate gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and high-purity germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10–90gU/L of natural uranyl nitrate are presented. A range of gamma-ray lines is examined, including attenuation for transmission measurement of density and concentration. It was determined that transmission-corrected gamma-ray spectra provide a reliable way to monitor the 235U concentration of uranyl nitrate solution in transfer pipes in NUCPs. Furthermore, existing predictive and analysis methods are adequate to design and realize practical designs. The 137Cs transmission source employed in this work is viable but not optimal for 235U densitometry determination. Validated simulations assessed the viability of 133Ba and 57Co as alternative densitometry sources. All three gamma-ray detectors are viable for monitoring natural uranium feed; although high-purity germanium is easiest to interpret, it is, however, the least attractive as an installation instrument. Overall, for monitoring throughput in a facility such as UNCLE, emulating the uranium concentration and pump speeds of the Springfields conversion facility in the United Kingdom, an uncertainty of less than 0.17% is required in order to detect the diversion of 1 SQ of uranyl nitrate through changes in uranium concentration over an accountancy period of one year with a detection probability of 50%. Although calibrated gamma-ray detection systems are capable of determining the concentration of uranium content in NUCPs, it is only in combination with verifiable operator declarations and supporting data, such as flow rate and enrichment, that safeguards conclusions can be drawn.

K-shell fluorescence yields and their uncertainties for use in hybrid K-edge densitometry

Hybrid K-edge densitometry (HKED) is a non-destructive analytical assay technique used to provide rapid determination of actinide concentration in tank solutions. Of special interest for HKED is the estimation, along with associated uncertainties, of the ratio of the flouresence yeilds, ωK, of uranium and plutonium. Limited experimental data for ωK(Z) as a function of atomic number, Z, exist and the data are subject to experimental uncertainty. Previous studies have provided values for ωK(Z) with uncertainty estimates but have not included covariance information. We use a phenomenological model with a bootstrapping method to generate the ratio ωK(94)/ωK(92) and associated uncertainty.

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel—Design concept and experimental demonstration

This paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. This paper describes the initial feasibility demonstration of the CIPN instrument, which involved measurements of four pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument׳s response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. These features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.

Development of PRIDE UNDA for safeguards application

In order to enhance the safeguardability of a pyroprocessing facility, the Korea Atomic Energy Research Institute (KAERI) has been endeavoring to develop more efficient and effective safeguards technologies for nuclear material accountancy (NMA), process monitoring, and containment and surveillance (C/S). NMA has two components: destructive analysis (DA) and non-destructive assay (NDA). Although DA is more accurate, it is typically time-consuming and cost-intensive. NDA, on the other hand, can provide reasonable accuracy on a real-time or near-real-time basis, which maximizes the utilization efficiency of a facility. In this study, the PRIDE (PyRoprocessing Integrated inactive DEmonstration) UNDA (unified non-destructive assay) was developed for testing NDA techniques at PRIDE, a demonstration facility within KAERI for integrated pyroprocessing using depleted uranium and surrogate materials. Each component of the PRIDE UNDA (i.e., neutron, gamma-ray, and mass measurement systems) was characterized and calibrated using calibration sources and standard weights as well as nuclear material used in the facility (depleted uranium). It is expected that in the near future, the PRIDE UNDA will be installed and tested with various types of process materials.

Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of 239Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3eV resonance of 239Pu. In addition, the measurement of the fast neutron flux is foreseen.This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements.Results from Monte Carlo simulations showed that ranges between 0.1–0.3mm and 0.5–1.0mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a 239Pu fission chamber is better suited to measure neutrons close to the 0.3eV resonance and it has the highest sensitivity to 239Pu, in comparison with a 235U fission chamber, with a 3He proportional counter, and with a 10B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the 239Pu and 235U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the 3He and 10B proportional counters to increase the sensitivity to 239Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies.

Integrated Nondestructive Assay Systems to Estimate Plutonium in Spent Fuel Assemblies

An integrated nondestructive assay (NDA) system combining active (neutron generator) and passive neutron detection and passive gamma (PG) detection is being analyzed in order to estimate the amount of plutonium, verify initial enrichment, burnup, and cooling time, and detect partial defects in a spent fuel assembly (SFA). Active signals are measured using the differential die-away (DDA), delayed neutron (DN), and delayed gamma (DG) techniques. Passive signals are measured using total neutron (TN) counts and both gross and spectral resolved gamma counts. To quantify how a system of several NDA techniques is expected to perform, all of the relevant NDA techniques listed above were simulated as a function of various reactor conditions such as initial enrichment, burnup, cooling time, assembly shuffling pattern, reactor operating conditions (including temperature, pressure, and the presence of burnable poisons) by simulating the NDA response for five sets of light water reactor assemblies. This paper compares the performance of several exploratory model-fitting options (including neural networks, adaptive regression with splines, iterative bias reduction smoothing, projection pursuit regression, and regression with quadratic terms and interaction terms) to relate data simulated with measurement and model error effects from various subsets of the NDA techniques to the total Pu mass. Isotope masses for SFAs and expected detector responses (DRs) for several NDA techniques are simulated using MCNP, and the DRs become inputs to the fitting process. Such responses include eight signals from DDA, one from DN, one from TN, and up to seven from PG; the DG signal will be examined separately. Results are summarized using the root-mean-squared estimation error for plutonium mass in held-out subsets of the data for a range of model and measurement error variances. Different simulation assumptions lead to different spent fuel libraries relating DRs to Pu mass. Some results for training with one library and testing with another library are also given.

Design optimization of radiation shielding structure for lead slowing-down spectrometer system

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux (>1012 n/cm2·s) neutron source comprised of a high-energy (30 MeV)/high-current (∼2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h), a few shielding materials [high-density polyethylene (HDPE)–Borax, B4C, and Li2CO3] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

Neutron Spectrometry for &lt;formula formulatype="inline"&gt;&lt;tex Notation="TeX"&gt;${\rm UF}_6$&lt;/tex&gt;&lt;/formula&gt; Enrichment Verification in Storage Cylinders

Verification of declared ${\rm UF}_6$ enrichment and mass in storage cylinders is of great interest in nuclear material nonproliferation. Nondestructive assay (NDA) techniques are commonly used for safeguards inspections to ensure accountancy of declared nuclear materials. Common NDA techniques used include gamma-ray spectrometry and both passive and active neutron measurements. In the present study, neutron spectrometry was investigated for verification of ${\rm UF}_6$ enrichment in 30B storage cylinders based on an unattended and passive measurement approach. MCNP5 and Geant4 simulated neutron spectra, for selected ${\rm UF}_6$ enrichments and filling profiles, were used in the investigation. The simulated neutron spectra were analyzed using principal component analysis (PCA). The PCA technique is a well-established technique and has a wide area of application including feature analysis, outlier detection, and gamma-ray spectral analysis. Results obtained demonstrate that neutron spectrometry supported by spectral feature analysis has potential for assaying ${\rm UF}_6$ enrichment in storage cylinders. Results from the present study also showed that difficulties associated with the ${\rm UF}_6$ filling profile and observed in other unattended passive neutron measurements can possibly be overcome using the approach presented.

Total neutron emission generation and characterization for a Next Generation Safeguards Initiative spent fuel library

In March of 2009 the Next Generation Safeguards Initiative of the U.S. Department of Energy began a nominal five year spent fuel research effort with the goal of: (1) quantifying plutonium content in spent nuclear fuel through the use of non-destructive assay (NDA) techniques; (2) quantifying the capability of these NDA techniques to quantify burnup, cooling time, and initial enrichment, as well as detect pin diversions, for spent fuel assemblies. These NDA techniques were first scoped in computation space in order to understand detection limitations and down select for further testing. Multiple Monte Carlo based spent fuel libraries (SFL's) were developed to be used as source terms for assessing the detection limitations of each of these techniques. The characterization of the SFL's provide an analysis of different burn strategies used for spent fuel, functional fits for state point analysis, and the components of the neutron signal.This paper details both a methodology for generating the spatially dependent gross neutron emission (GNE) as well as a characterization of the GNE for a particular NGSI SFL for two different shuffle patterns. The GNE was characterized by examining three major components: (1) the total neutron emission; (2) the (α, n) and spontaneous fission emissions; and (3) the main contributors to the spontaneous fission and (α, n) emission signals. The total neutron emission was characterized by looking at two different shuffle patterns and using three different pin analysis techniques. The pin analysis techniques used were: (1) sum all the pins in the assembly; (2) create four pin zones to be summed; and (3) examine specific pins. These characterizations provide a large set of data with integral values (cooling time vs emission and burnup vs emission) and derivative values (isotopic percent composition) with visual depiction and detailed explanation. The characterizations also provide a comparison of the effects of shuffle schemes on the neutron emission signal for two different shuffle patterns. This data can be used in the development of time and burnup dependent normalized functions to describe the trends displayed in the data. This study also shows that the information which can be obtained from the GNE depends upon how the information is grouped and analyzed. Summing all the pins of the assembly or assuming a uniform burnup results in the loss of valuable neutron emission information. By splitting the pins into zones, trends such as the total emissions across the assembly become evident. By evaluating individual pins trends such as the effect of water rods on the neutron emission become evident. This type of analysis can be of great use to safeguards. Evaluating the GNE on a pin by pin basis leads to a much clearer picture of the constituents of the assembly.

Investigation of the Self-Interrogation Neutron Resonance Densitometry applied to spent fuel using Monte Carlo simulations

Self-Interrogation Neutron Resonance Densitometry (SINRD) is a Non-Destructive Assay technique that has the potential to directly quantify 239Pu in spent fuel. SINRD is based on the detection of the neutron flux attenuation by 239Pu around the 0.3eV neutron resonance, here indicated as the SINRD signature. An effective deployment of the SINRD technique relies on demonstrating that the 239Pu content in spent fuel can be correlated with the SINRD signature.Previous work from Los Alamos National Laboratory proposed to place a set of fission chambers on one side of the fuel assembly and to detect the passive neutron emission. Since it is expected that in water the signature degrades after a few millimeters from the fuel assembly, a different measurement approach is proposed in our work, namely to insert a set of fission chambers in the guide tubes of the fuel assembly.SCK•CEN conducted a research based on Monte Carlo simulations to investigate the phenomena affecting the SINRD signature. First, we studied the impact of measurement conditions on the energy distribution of the neutron flux at the measurement position. Then, we determined the sensitivity of the SINRD signature to 239Pu content. In the final part of our research, we assessed the expected individual impact that other nuclides have on the SINRD signature.We concluded that a better sensitivity is expected when the fuel assembly is in air, surrounded by a slab of polyethylene. The obtained results also indicated that the ratio between the neutron flux in the fast region and the neutron flux around the 0.3eV resonance is more strongly related to the 239Pu content than other ratios. The impact of 235U on the SINRD signature was found to be significant, while limited effects were due to other nuclides.

Quantitative NDA Measurements of Multiactinide Oxide Fuels

As new reprocessing techniques and fuel forms are developed, the ability of inspection agencies and facility operators to measure powders containing several actinides becomes increasingly necessary. Neutrons emitted from induced and spontaneous fission of different nuclides are very similar, making it difficult to measure these powders with nondestructive assay techniques. To measure the powders, a neutron multiplicity technique based on first-principle methods was developed to exploit isotope-specific nuclear properties, such as energy-dependent fission cross sections and neutron-induced fission multiplicity. This technique was tested through measurements using an epithermal neutron multiplicity counter with two different interrogation (α,n) sources and varying plutonium materials. To complement these measurements, extensive Monte Carlo N-Particle eXtended (MCNPX) simulations were performed for each measured sample, as well as samples that were not available to measure. The primary application of this first-principle technique is the measurement of materials containing uranium, neptunium, plutonium, and americium. This technique still has several challenges that need to be overcome, the largest of these being the ability to produce results with acceptably small uncertainties.

Validation for application of the Monte Carlo simulation code for 235U mass content verification for large size samples of nuclear materials

In this work, a new semi- absolute non-destructive assay technique has been developed to verify the mass content of 235U in the large sizes nuclear material samples of different enrichment through combination of experimental measurements and Mont Carlo calculations (version MCNP5). A good agreement was found between the calculated and declared values of the mass content of 235U of uranium oxide (UO2) samples. The results obtained from Mont Carlo calculations showed that the semi-absolute technique can be used with sufficient reliability to verify the uranium mass content in the large sizes nuclear material samples of different enrichment.

Development of an analytical theory to describe the PNAR and CIPN nondestructive assay techniques

This paper develops an analytical theory to describe how neutron albedo (reflection) increases the multiplication of neutrons by a used fuel assembly. With this theory, the two nondestructive assay (NDA) techniques of Passive Neutron Albedo Reactivity (PNAR) and Californium-252 Interrogation with Prompt Neutron Detection (CIPN) can be compared directly. Specifically, the theory derives expressions for the PNAR and CIPN metrics in terms of the physical properties of the used fuel assembly, such as the neutron multiplications and fate probabilities. The theory thus clarifies the interpretation of these two NDA techniques and suggests ways to improve both the design of the NDA instruments and the algorithms for analyzing the measurement results.

Characterization and performance evaluation of a new passive neutron albedo reactivity counter for safeguards measurements

A prototype 3He-based Passive Neutron Albedo Reactivity (PNAR) counter was developed and tested at Los Alamos National Laboratory (LANL) in collaboration with the Korea Atomic Energy Research Institute (KAERI) to measure the fissile content in electrochemical recycling (ER) product materials. The counter consists of 16 3He cylindrical gas-filled proportional counters at 4 atm of pressure embedded in high-density polyethylene. In this work, experimental measurements were performed at LANL to characterize the performance of the PNAR counter using surrogate materials for the uranium metal ingot. The purpose of these experiments was to: 1) measure the operating and calibration parameters of the PNAR counter (e.g. efficiency profiles, coincidence gate fractions, die-away time) and 2) evaluate the accuracy and sensitivity of the PNAR method and the time correlated induced fission (TCIF) method for quantifying the 235U mass in PWR fresh LEU fuel rods and Materials Testing Reactor (MTR) HEU fuel plates. A small 244Cm reference source (13,373 n/s) was placed in the center of the fuel rods and fuel plates to simulate spontaneous fission from sub-ppm (parts per million) levels of Cm contamination in the U ingot. In order to compare the relative accuracy of the PNAR and TCIF methods for quantifying 235U mass, calibration curves were generated for the net doubles rate and the doubles Cd ratio using the Deming software. The results from this experiment will be used to obtain a better understanding of the sensitivity of the PNAR and TCIF methods for samples with low neutron multiplication. Furthermore, this experimental measurement data will also help inform safeguards research and development (R&D) efforts on the viability of nondestructive assay (NDA) techniques and detector designs for quantifying fissile content in ER product materials. Future work will include performing measurements with the PNAR counter on small samples of U/TRU materials.

Development of a neutron measurement system in unified non-destructive assay for the PRIDE facility

The Korea Atomic Energy Research Institute (KAERI) has made an effort to develop pyroprocessing technology to resolve an on-going problem in Korea, i.e., the management of spent nuclear fuels. To this end, a test-bed facility for pyroprocessing, called PRIDE (PyRoprocessing Integrated inactive DEmonstration facility), is being constructed at KAERI. The main objective of PRIDE is to evaluate the performance of the unit processes, remote operation, maintenance, and proliferation resistance. In addition, integrating all unit processes into a one-step process is also one of the main goals. PRIDE can also provide a good opportunity to test safeguards instrumentations for a pyroprocessing facility such as nuclear material accounting devices, surveillance systems, radiation monitoring systems, and process monitoring systems. In the present study, a non-destructive assay (NDA) system for the testing of nuclear material accountancy of PRIDE was designed by integrating three different NDA techniques, i.e., neutron, gamma-ray, and mass measurements. The developed neutron detection module consists of 56 3He tubes and 16 AMPTEK A111 signal processing circuits. The amplifiers were matched in terms of the gain and showed good uniformity after a gain-matching procedure (%RSD=0.37%). The axial and the radial efficiency distributions within the cavity were then measured using a 252Cf neutron source and were compared with the MCNPX calculation results. The measured efficiency distributions showed excellent agreement with the calculations, which confirmed the accuracy of the MCNPX model of the system.