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A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation
Abstract and Figures
The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the ²³⁵U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the ²³⁸U atoms in the fuel rods to self-interrogate the ²³⁵U mass. The innovation for the new approach is that the ²³⁸U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the ²³⁵U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the ²³⁸U SF (2.07) and the ²³⁵U IF (2.44) for a fraction of the IF reactions. Whereas, the ²³⁸U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the ²³⁵U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).
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This chapter presents a description of most of the instruments that are currently in use for the measurement of plutonium and uranium using passive methods (without an external source). This includes the acquisition electronics as well as Singles counting methods, coincidence counting methods and multiplicity counting methods. The Singles counting applications include the measurement of waste and curium bearing materials. The coincidence counting applications include bulk plutonium, bulk uranium, waste and holdup measurements and fresh fuel assemblies. The multiplicity application description includes advantages and disadvantages and multiplicity detector design. There is also a description of some non-3He systems. The chapter concludes with a description of additional concepts: neutron imagers, list-mode data analysis, distributed source term analysis, unattended monitoring and MCNP modeling for detector design.
A reliable 235U enrichment uniformity detection system based on a compact D-D neutron generator is developed to detect the 235U enrichment uniformity of different fuel elements in the same nuclear fuel rod. The high-yield compact D-D neutron generator provides 2.45 MeV D-D neutrons, decelerated by a moderator to thermal neutrons or epithermal neutrons, thereby inducing 235U fission to produce highly excited state fission fragments that undergo de-excitation via γ-ray emission. The system detects the 235U enrichment uniformity of a nuclear fuel rod by measuring γ-rays and establishing a relationship between the γ-ray count rate and 235U enrichment in nuclear fuel. The proposed system yields a confidence probability of 99.99% for a relative 235U enrichment deviation of 10% and a neutron yield of 5 × 108 n/s, and the detection accuracy increases with increasing neutron yield. Furthermore, the developed system can satisfy quality control requirements for nuclear fuel production to promote the safe development of nuclear power.
One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different ²³⁵U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.
Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies.
This paper presents the first application of a new technique for the measurement of the 235U content in fresh fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by multiplicity counting. The doubles and triples rates are enhanced by having the trigger events from both the 252Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission (ν) of the 252Cf source is 3.76 and the induced fission ν for 235U is 2.44, so the combined ν is ∼5.2 with one neutron removed by the fission reaction. This high effective ν significantly increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252Cf have been minimized by neutron shielding between the source and the detector. This method of active neutron interrogation has been applied to the measurement of fresh pressurized water reactor (PWR) fuel assemblies. The neutron uranium collar (UNCL) that is routinely used for 235U verification in PWR reactor fuel assemblies is used to compare the TCIF method with the typically used AmLi neutron interrogation source. This paper presents both the experimental verification of the TCIF method for a PWR mockup assembly and the MCMPX simulations to optimize the detector geometry.
Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.
The neutron coincidence collar is used to verify the uranium content in light water reactor fuel assemblies. An AmLi neutron source actively interrogates the fuel assembly to measure the ²³âµU content and the ²³â¸U content can be verified from a passive neutron coincidence measurement. This report gives the collar calibration data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies both with and without cadmium liners. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and various fuel assembly sizes. The data were collected using the Los Alamos BWR and PWR test assemblies as well as fuel assemblies from several fuel fabrication facilities. 11 refs., 15 figs., 14 tabs.
This document is intended to serve as a comprehensive applications guide to passive
neutron multiplicity counting, a new nondestructive assay (NDA) technique developed over the
past ten years. The document describes the principles of multiplicity counter design, electronics,
and mathematics. Existing counters in Department of Energy (DOE) facilities are surveyed, and
their operating requirements and procedures and defined. Current applications to plutonium
material types found in DOE facilities are described, and estimates of the expected assay precision
and bias are given. Lastly, guidelines for multiplicity counter selection and procurement are
summarized. The document also includes a detailed collection of references on passive neutron
coincidence and multiplicity publications over the last ten to fifteen years.
Advanced Neutron Detection Technology Rodeo
- Belian Anthony
Belian Anthony, et al., Advanced Neutron Detection Technology Rodeo, presented
at ESARDA 2017 Annual Meeting, Dusseldorf, Germany, May, 2017.
Liquid Scintillator-Based Fast Neutron Coincidence Counter for Fresh Nuclear Fuel Measurements
- T H Lee
- A Tomanin
- J Beaumont
T.H. Lee, A. Tomanin, J. Beaumont, Liquid Scintillator-Based Fast Neutron Coincidence Counter for Fresh Nuclear Fuel Measurements, American Nuclear Society,
Santa Fe, 2016 Advances in Nuclear Nonproliferation Technology and Policy Conference.
Passive Nondestructive Assay of Nuclear Materials Washington DC, United States Nuclear Regulatory Commission
- D Reilly
- N Ensslin
- H A Smith
- S Kreiner
D. Reilly, N. Ensslin, H.A. Smith, Jr., and S. Kreiner, Eds., Passive Nondestructive
Assay of Nuclear Materials Washington DC, United States Nuclear Regulatory
Commission, NUREG/CR-5550, LA-UR-90-732,1991.
Correlated Fission Multiplicity Model Verification Efforts in MCNP6
- Rising Michael Evan
Rising Michael Evan, Correlated Fission Multiplicity Model Verification Efforts in
MCNP6, American Nuclear Society Advances in Nuclear Nonproliferation Technology and Policy Conference, 2016-09-25/2016-09-30, Santa Fe, New Mexico, United
States, LA-UR-16-23341, 2016.
High-Dose Neutron Detector Development for Measuring Alternative Fuel Cycle Materials, LA-UR-17-23325
- D Henzlova
- H O Menlove
D. Henzlova, H.O. Menlove, High-Dose Neutron Detector Development for Measuring Alternative Fuel Cycle Materials, LA-UR-17-23325, Proceedings of GLOBAL
2017, Seoul Korea September 24-29, 2017.