Neutron Generators for Spent Fuel Assay

The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel (SNF) assemblies with non-destructive assay (NDA). The 14 NDA techniques being studied include several that require an external neutron source: Delayed Neutrons (DN), Differential Die-Away (DDA), Delayed Gammas (DG), and Lead Slowing-Down Spectroscopy (LSDS). This report provides a survey of currently available neutron sources and their underlying technology that may be suitable for NDA of SNF assemblies. The neutron sources considered here fall into two broad categories. The term 'neutron generator' is commonly used for sealed devices that operate at relatively low acceleration voltages of less than 150 kV. Systems that employ an acceleration structure to produce ion beam energies from hundreds of keV to several MeV, and that are pumped down to vacuum during operation, rather than being sealed units, are usually referred to as 'accelerator-driven neutron sources.' Currently available neutron sources and future options are evaluated within the parameter space of the neutron generator/source requirements as currently understood and summarized in section 2. Applicable neutron source technologies are described in section 3. Commercially available neutron generators and other source options that could be made available in the near future with some further development and customization are discussed in sections 4 and 5, respectively. The pros and cons of the various options and possible ways forward are discussed in section 6. Selection of the best approach must take a number of parameters into account including cost, size, lifetime, and power consumption, as well as neutron flux, neutron energy spectrum, and pulse structure that satisfy the requirements of the NDA instrument to be built

Using Nuclear Resonance Fluorescence for Nondestructive Isotopic Analysis

Nuclear resonance fluorescence (NRF) has been studied as one of the nondestructive analysis (NDA) techniques currently being investigated by a multi-laboratory collaboration for the determination of Pu mass in spent fuel. In NRF measurements specific isotopes are identified by their characteristic lines in recorded gamma spectra. The concentration of an isotope in a material can be determined from measured NRF signal intensities if NRF cross sections and assay geometries are known. The potential of NRF to quantify isotopic content and Pu mass in spent fuel has been studied. The addition of NRF data to MCNPX and an improved treatment of the elastic photon scattering at backward angles has enabled us to more accurately simulate NRF measurements on spent fuel assemblies. Using assembly models from the spent fuel assembly library generated at LANL, NRF measurements are simulated to find the best measurement configurations, and to determine measurement sensitivities and times, and photon source and gamma detector requirements. A first proof-of-principal measurement on a mock-up assembly with a bremsstrahlung photon source demonstrated isotopic sensitivity to approximately 1% limited by counting statistics. Data collection rates are likely a limiting factor of NRF-based measurements of fuel assemblies but new technological advances may lead to drastic improvements

Nuclear Resonance Fluorescence for Safeguards Applications

In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of {gamma} rays with specific energies that are characteristic of the emitting isotope. The promise of NRF as a non-destructive analysis technique (NDA) in safeguards applications lies in its potential to directly quantify a specific isotope in an assay target without the need for unfolding the combined responses of several fissile isotopes as often required by other NDA methods. The use of NRF for detection of sensitive nuclear materials and other contraband has been researched in the past. In the safeguards applications considered here one has to go beyond mere detection and precisely quantify the isotopic content, a challenge that is discussed throughout this report. Basic NRF measurement methods, instrumentation, and the analytical calculation of NRF signal strengths are described in Section 2. Well understood modeling and simulation tools are needed for assessing the potential of NRF for safeguards and for designing measurement systems. All our simulations were performed with the radiation transport code MCNPX, a code that is widely used in the safeguards community. Our initial studies showed that MCNPX grossly underestimated the elastically scattered background at backwards angles due to an incorrect treatment of Rayleigh scattering. While new, corrected calculations based on ENDF form factors showed much better agreement with experimental data for the elastic scattering of photons on an uranium target, the elastic backscatter is still not rigorously treated. Photonuclear scattering processes (nuclear Thomson, Delbruck and Giant Dipole Resonance scattering), which are expected to play an important role at higher energies, are not yet included. These missing elastic scattering contributions were studied and their importance evaluated evaluated against data found in the literature as discussed in Section 3. A transmission experiment was performed in September 2009 to test and demonstrate the applicability of the method to the quantitative measurement of an isotope of interest embedded in a thick target. The experiment, data analysis, and results are described in Section 4. The broad goal of our NRF studies is to assess the potential of the technique in safeguards applications. Three examples are analyzed in Section 5: the isotopic assay of spent nuclear fuel (SNF), the measurement of {sup 235}U enrichment in UF{sub 6} cylinders, and the determination of {sup 239}Pu in mixed oxide (MOX) fuel. The study of NRF for the assay of SNF assemblies was supported by the Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy as part of a large multi-lab/university effort to quantify the plutonium (Pu) mass in spent nuclear fuel assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. NRF is one of 14 NDA techniques being researched. The methodology for performing and analyzing quantitative NRF measurements was developed for determining Pu mass in SNF and is extensively discussed in this report. The same methodology was applied to the assessment of NRF for the measurement of {sup 235}U enrichment and the determination of {sup 239}Pu in MOX fuel. The analysis centers on determining suitable NRF measurement methods, measurement capabilities that could be realized with currently available instrumentation, and photon source and detector requirements for achieving useful NDA capabilities

Nuclear Resonance Fluorescence for Safeguards Applications

In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of {gamma} rays with specific energies that are characteristic of the emitting isotope. The promise of NRF as a non-destructive analysis technique (NDA) in safeguards applications lies in its potential to directly quantify a specific isotope in an assay target without the need for unfolding the combined responses of several fissile isotopes as often required by other NDA methods. The use of NRF for detection of sensitive nuclear materials and other contraband has been researched in the past. In the safeguards applications considered here one has to go beyond mere detection and precisely quantify the isotopic content, a challenge that is discussed throughout this report. Basic NRF measurement methods, instrumentation, and the analytical calculation of NRF signal strengths are described in Section 2. Well understood modeling and simulation tools are needed for assessing the potential of NRF for safeguards and for designing measurement systems. All our simulations were performed with the radiation transport code MCNPX, a code that is widely used in the safeguards community. Our initial studies showed that MCNPX grossly underestimated the elastically scattered background at backwards angles due to an incorrect treatment of Rayleigh scattering. While new, corrected calculations based on ENDF form factors showed much better agreement with experimental data for the elastic scattering of photons on an uranium target, the elastic backscatter is still not rigorously treated. Photonuclear scattering processes (nuclear Thomson, Delbruck and Giant Dipole Resonance scattering), which are expected to play an important role at higher energies, are not yet included. These missing elastic scattering contributions were studied and their importance evaluated evaluated against data found in the literature as discussed in Section 3. A transmission experiment was performed in September 2009 to test and demonstrate the applicability of the method to the quantitative measurement of an isotope of interest embedded in a thick target. The experiment, data analysis, and results are described in Section 4. The broad goal of our NRF studies is to assess the potential of the technique in safeguards applications. Three examples are analyzed in Section 5: the isotopic assay of spent nuclear fuel (SNF), the measurement of {sup 235}U enrichment in UF{sub 6} cylinders, and the determination of {sup 239}Pu in mixed oxide (MOX) fuel. The study of NRF for the assay of SNF assemblies was supported by the Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy as part of a large multi-lab/university effort to quantify the plutonium (Pu) mass in spent nuclear fuel assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. NRF is one of 14 NDA techniques being researched. The methodology for performing and analyzing quantitative NRF measurements was developed for determining Pu mass in SNF and is extensively discussed in this report. The same methodology was applied to the assessment of NRF for the measurement of {sup 235}U enrichment and the determination of {sup 239}Pu in MOX fuel. The analysis centers on determining suitable NRF measurement methods, measurement capabilities that could be realized with currently available instrumentation, and photon source and detector requirements for achieving useful NDA capabilities

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size, development time, nor concerns related to the use of Pu in measurement systems. This report discusses basic NRF measurement concepts, i.e., backscatter and transmission methods, and photon source and {gamma}-ray detector options in Section 2. An analytical model for calculating NRF signal strengths is presented in Section 3 together with enhancements to the MCNPX code and descriptions of modeling techniques that were drawn upon in the following sections. Making extensive use of the model and MCNPX simulations, the capabilities of the backscatter and transmission methods based on bremsstrahlung or quasi-monoenergetic photon sources were analyzed as described in Sections 4 and 5. A recent transmission experiment is reported on in Appendix A. While this experiment was not directly part of this project, its results provide an important reference point for our analytical estimates and MCNPX simulations. Used fuel radioactivity calculations, the enhancements to the MCNPX code, and details of the MCNPX simulations are documented in the other appendices

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size, development time, nor concerns related to the use of Pu in measurement systems. This report discusses basic NRF measurement concepts, i.e., backscatter and transmission methods, and photon source and {gamma}-ray detector options in Section 2. An analytical model for calculating NRF signal strengths is presented in Section 3 together with enhancements to the MCNPX code and descriptions of modeling techniques that were drawn upon in the following sections. Making extensive use of the model and MCNPX simulations, the capabilities of the backscatter and transmission methods based on bremsstrahlung or quasi-monoenergetic photon sources were analyzed as described in Sections 4 and 5. A recent transmission experiment is reported on in Appendix A. While this experiment was not directly part of this project, its results provide an important reference point for our analytical estimates and MCNPX simulations. Used fuel radioactivity calculations, the enhancements to the MCNPX code, and details of the MCNPX simulations are documented in the other appendices

Coaxial Mono-Energetic Gamma Generator for Active Interrogation

Compact mono-energetic photon sources are sought for active interrogation systems to detect shielded special nuclear materials in, for example, cargo containers, trucks and other vehicles. A prototype gamma interrogation source has been designed and built that utilizes the 11B(p,gamma)12C reaction to produce 12 MeV gamma-rays which are near the peak of the photofission cross section. In particular, the 11B(p,gamma)12C resonance at 163 kV allows the production of gammas at low proton acceleration voltages, thus keeping the design of a gamma generator comparatively small and simple. A coaxial design has been adopted with a toroidal-shaped plasma chamber surrounding a cylindrical gamma production target. The plasma discharge is driven by a 2 MHz rf-power supply (capable up to 50 kW) using a circular rf-antenna. Permanent magnets embedded in the walls of the plasma chamber generate a multi-cusp field that confines the plasma and allows higher plasma densities and lower gas pressures. About 100 proton beamlets are extracted through a slotted plasma electrode towards the target at the center of the device that is at a negative 180 kV. The target consists of LaB6 tiles that are brazed to a water-cooled cylindrical structure. The generator is designed to operate at 500 Hz with 20 mu s long pulses, and a 1percent duty factor by pulsing the ion source rf-power. A first-generation coaxial gamma source has been built for low duty factor experiments and testing

Coaxial Mono-Energetic Gamma Generator for Active Interrogation

Compact mono-energetic photon sources are sought for active interrogation systems to detect shielded special nuclear materials in, for example, cargo containers, trucks and other vehicles. A prototype gamma interrogation source has been designed and built that utilizes the 11B(p,gamma)12C reaction to produce 12 MeV gamma-rays which are near the peak of the photofission cross section. In particular, the 11B(p,gamma)12C resonance at 163 kV allows the production of gammas at low proton acceleration voltages, thus keeping the design of a gamma generator comparatively small and simple. A coaxial design has been adopted with a toroidal-shaped plasma chamber surrounding a cylindrical gamma production target. The plasma discharge is driven by a 2 MHz rf-power supply (capable up to 50 kW) using a circular rf-antenna. Permanent magnets embedded in the walls of the plasma chamber generate a multi-cusp field that confines the plasma and allows higher plasma densities and lower gas pressures. About 100 proton beamlets are extracted through a slotted plasma electrode towards the target at the center of the device that is at a negative 180 kV. The target consists of LaB6 tiles that are brazed to a water-cooled cylindrical structure. The generator is designed to operate at 500 Hz with 20 mu s long pulses, and a 1percent duty factor by pulsing the ion source rf-power. A first-generation coaxial gamma source has been built for low duty factor experiments and testing