Neutron Resonance Densitometry Research Articles

SINRD validation experiments at the time-of-flight facility GELINA

Self-interrogation neutron resonance densitometry (SINRD) is a non-destructive analysis technique that can be used to quantify the amount of 239Pu in spent nuclear fuel. It is a passive method that relies on the detection of neutrons, which are emitted by the fuel. The amount of 239Pu is estimated from the ratio of the neutron intensity in the fast energy region and in a region close to the 0.296eV resonance of 239Pu. The neutron intensity in the resonance region is obtained from a detection system with a high sensitivity to 0.296eV neutrons. This can be realized by using two neutron detectors with 239Pu as convertor material. One of the detectors is covered by a thin Gd foil and the other by a thin Cd foil. The Gd and Cd foils are referred to as SINRD filters.An approach based on the measurement of a fuel assembly in air and surrounded by a slab of polyethylene was developed at SCK·CEN. This approach foresees the insertion of small neutron detectors in the guide tubes of the assembly, and optimisation studies of SINRD were based on Monte Carlo simulations. Experiments to support the results of such simulations were carried out at the time-of-flight facility GELINA of the Joint Research Centre (JRC) in Geel (Belgium). Transmission measurements were performed to verify the quality of the nuclear data that are used to define the optimum thickness of the SINRD filters. Results of self-indication measurements were used to confirm the basic principle of SINRD, that is, that the best results are obtained with a detector that has a high sensitivity to neutrons with energy close to the energy of a strong resonance of the material under investigation. The results of these experiments are presented in this work.

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.

Experiments at the GELINA facility for the validation of the self-indication neutron resonance densitometry technique

Self-Indication Neutron Resonance Densitometry (SINRD) is a passive non-destructive method that is being investigated to quantify the 239 Pu content in a spent fuel assembly. The technique relies on the energy dependence of total cross sections for neutron induced reaction. The cross sections show resonance structures that can be used to quantify the presence of materials in objects, e.g. the total cross-section of 239 Pu shows a strong resonance close to 0.3 eV. This resonance will cause a reduction of the number of neutrons emitted from spent fuel when 239 Pu is present. Hence such a reduction can be used to quantify the amount of 239 Pu present in the fuel. A neutron detector with a high sensitivity to neutrons in this energy region is used to enhance the sensitivity to 239 Pu. This principle is similar to self-indication cross section measurements. An appropriate detector can be realized by surrounding a 239 Pu-loaded fission chamber with appropriate neutron absorbing material. In this contribution experiments performed at the GELINA time-of-flight facility of the JRC at Geel (Belgium) to validate the simulations are discussed. The results confirm that the strongest sensitivity to the target material was achieved with the self-indication technique, highlighting the importance of using a 239 Pu fission chamber for the SINRD measurements.

LaBr3γ-ray spectrometer for detecting 10B in debris of melted nuclear fuel

Neutron resonance densitometry has been proposed as a nondestructive analytical method for quantifying special nuclear material (SNM) in the rock- and particle-like debris that is to be removed from the Fukushima Daiichi nuclear power plant. The method is based on neutron resonance transmission analysis (NRTA) and neutron resonance capture analysis combined with prompt-γ-ray analysis (NRCA/PGA). Although quantification of SNM will predominantly rely on NRTA, this will be hampered by the presence of strong neutron-absorbing matrix materials, in particular 10B. Results obtained with NRCA/PGA are used to improve the interpretation of NRTA data. Prompt γ rays originating from the 10B(n, αγ) reaction are used to assess the amount of 10B. The 478keV γ rays from 10B, however, need to be measured under a high-radiation environment, especially because of 137Cs. To meet this requirement, we developed a well-shaped γ-ray spectrometer consisting of one cylindrical and four rectangular-cuboid LaBr3 scintillators combined with a fast data-acquisition system. Furthermore, to improve the gain stability of the main detector, a special high-voltage divider was developed. Because of the reduction in gain shift, a 3.8% resolution at 662keV was obtained for long-term measurements. By using the data-acquisition system, which consists of eight 250MHz digitizers, input signals of over 500kHz per channel were recorded. The work reported herein demonstrates that, with such a spectrometer, the impact of the Compton edge of 662keV γ rays from 137Cs is significantly reduced, which allows the 10B amount to be determined with greater sensitivity.

Development of Neutron Resonance Densitometry as a Technique of Material Accountancy for Nuclear Materials with Complicated Shapes and Components

東京電力福島第一原子力発電所における冷却材喪失過酷事故により，1-3号炉では炉心溶融が起きたと考えられている。溶融燃料(デブリ)は，一定の冷却期間をおいて取り出される計画となっている。通常の原子炉では，燃料集合体を単位とした核物質計量管理を行っているが，今回のように燃料集合体が溶融した場合，取り出した核物質量を何らかの測定を行った上で計量管理することが求められる可能性がある。そうした中，粒子状溶融燃料中の核物質量を定量可能とする非破壊測定技術として中性子共鳴濃度分析法を考案し，平成24年度より3年間，文部科学省による核不拡散・核セキュリティ分野における新規課題として技術開発を進めてきた。本解説では，開発した技術を概説するとともに，国際原子力機関等の専門家が集まるワークショップで行った技術実証試験について報告する。

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.

Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies.

Generalized analysis method for neutron resonance transmission analysis

Neutron resonance densitometry (NRD) is a non-destructive analysis method, which can be applied to quantify special nuclear materials (SNM) in small particle-like debris of melted fuel that are formed in severe accidents of nuclear reactors such as the Fukushima Daiichi nuclear power plants. NRD uses neutron resonance transmission analysis (NRTA) to quantify SNM and neutron resonance capture analysis (NRCA) to identify matrix materials and impurities. To apply NRD for the characterization of arbitrary-shaped thick materials, a generalized method for the analysis of NRTA data has been developed. The method has been applied on data resulting from transmission through thick samples with an irregular shape and an areal density of SNM up to 0.253 at/b (≈100 g/cm2). The investigation shows that NRD can be used to quantify SNM with a high accuracy not only in inhomogeneous samples made of particle-like debris but also in samples made of large rocks with an irregular shape by applying the generalized analysis method for NRTA.

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.

Uncertainty assessment of neutron resonance transmission analysis using the linear absorption model

Neutron resonance densitometry (NRD) has been proposed to quantify nuclear materials in particle-like debris of melted spent nuclear fuel formed in severe nuclear reactor accidents. The NRD is a hybrid technique of the neutron resonance transmission analysis and the neutron resonance capture analysis. We have studied the statistical uncertainties of the neutron resonance transmission analysis by using the linear absorption model. Particularly, the effects of impurities and sample thickness on the uncertainties were examined.

Neutron Resonance Densitometry for Particle-like Debris of Melted Fuel

Neutron Resonance Densitometry (NRD) is proposed for the quantification of nuclear materials in particle-like debris of melted fuel from the reactors of the Fukushima Daiichi nuclear power plant. The method is based on a combination of neutron resonance transmission analysis (NRTA) and neutron resonance capture analysis (NRCA). It uses the neutron time-of-flight (TOF) technique with a pulsed white neutron source and a neutron flight path as short as 5 m. The spectrometer for NRCA is made of LaBr3(Ce) detectors. The achievable uncertainty due to only counting statistics is less than 1 % to determine Pu and U isotopes.

A Monte Carlo simulation to study a design of a gamma-ray detector for neutron resonance densitometry

Neutron resonance densitometry (NRD) has been proposed to quantify nuclear materials in melted fuel (MF) that will be removed from the Fukushima Daiichi nuclear power plant. The problem is complex due to the expected presence of strong neutron absorbing impurities such as 10B and high radiation field that is mainly caused by 137Cs. To identify the impurities under the high radiation field, NRD is based on a combination of neutron resonance transmission analysis (NRTA) and neutron resonance capture analysis (NRCA). We investigated with Geant4 the performance of a gamma-ray detector for NRCA in NRD. The gamma-ray detector has a well shape, consisting of cylindrical and tube type LaBr3 scintillators. We show how it measures 478keV gamma rays derived from 10B(n, αγ) reaction in MF under a high 137Cs-radiation environment. It was found that the gamma-ray detector was able to well suppress the Compton edge of 662-keV gamma rays of 137Cs and had a high peak-to-Compton continuum ratio, by using the tube type scintillator as a back-catcher detector. Then, we demonstrate that with this ability, detection of 478-keV gamma rays from 10B is accomplished in realistic measuring time.

パルス中性子源を利用した中性子共鳴濃度分析法の開発:(3)含有ボロンの中性子透過スペクトルへの影響評価

パルス中性子源を利用した中性子共鳴濃度分析法の開発:(3)含有ボロンの中性子透過スペクトルへの影響評価

Performance assessment of self-interrogation neutron resonance densitometry for spent nuclear fuel assay

Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the 239Pu and 235U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of 239Pu and 235U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that 235U content has a major impact on the SINRD signal in addition to the 239Pu content. Plutonium-241 and 241Am are the two main nuclides responsible for the variation in the count rate ratio with cooling time. In short, this work provides insights into some of the main factors that affect the performance of SINRD, and it should help improve the hardware design and the algorithm used to interpret the signal for the SINRD technique. In addition, the modeling and simulation techniques used in this work can be easily adopted for analysis of other NDA systems, especially when complex systems like spent nuclear fuel are involved. These studies were conducted at Los Alamos National Laboratory.

Development of Self-Interrogation Neutron Resonance Densitometry to Improve Detection of Possible Diversions for PWR Spent Fuel Assemblies

A new nondestructive assay technique called self-interrogation neutron resonance densitometry (SINRD) is currently being developed at Los Alamos National Laboratory to improve existing nuclear safeguards and material accountability measurements for light water reactor fuel assemblies. The viability of using SINRD to improve the detection of possible diversion scenarios for pressurized water reactor 17 × 17 spent low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The following capabilities were assessed: (a) verification of the burnup of a spent fuel assembly, (b) ability to distinguish fresh and one-cycle spent MOX fuel from three- and four-cycle spent LEU fuel, and (c) sensitivity and penetrability to the removal of fuel pins. SINRD utilizes 244Cm spontaneous-fission neutrons to self-interrogate the spent fuel pins. The amount of resonance absorption of these neutrons in the fuel can be quantified using a set of fission chambers (FCs) placed adjacent to the assembly. The sensitivity of SINRD is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in the FC. SINRD requires calibration with a reference assembly of similar geometry in a similar measurement configuration with the same surrounding moderator as the spent fuel assemblies. However, this densitometry method uses ratios of different detectors so that several systematic errors related to calibration and positioning cancel in the ratios.

パルス中性子源を利用した中性子共鳴濃度分析法の開発:(1) 分析原理及び開発計画

パルス中性子源を利用した中性子共鳴濃度分析法の開発:(1) 分析原理及び開発計画

Development of Self-Interrogation Neutron Resonance Densitometry to Quantify the Fissile Content in PWR Spent LEU and MOX Assemblies

A new nondestructive assay technique called self-interrogation neutron resonance densitometry (SINRD) is currently being developed at Los Alamos National Laboratory to improve existing nuclear safeguards and material accountability measurements for light water reactor fuel assemblies. The viability of using SINRD to quantify the fissile content (235U and 239Pu) in pressurized water reactor 17 × 17 spent low-enriched uranium and mixed-oxide fuel assemblies in water was investigated via Monte Carlo N-particle extended transport code simulations. SINRD utilizes 244Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be quantified using 235U and 239Pu fission chambers placed adjacent to the assembly. The sensitivity of this technique is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. SINRD requires calibration with a reference assembly of similar geometry. However, this densitometry method uses ratios of different fission chambers so that most systematic errors related to calibration and positioning cancel in the ratios.

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.