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This paper is an examination of the effect of an axially varying fluid temperature on the measurement of $\Delta {T}({z})$ , the difference in the temperature between the optical fiber within the thermal mass and the optical fiber within the capillary tube that is attached to the outer sheath of an Optical Fiber Based Gamma Calorimeter (OFBGC). Our analysis indicates that for an OFBGC with the general design of the OFBGC that was built and tested in the Ohio State University Research Reactor (OSURR), a sinusoidal variation with spatial frequency ${k}_{B}$ in the bulk fluid temperature ${T}_{B}$ along the length of the OFBGC will create a sinusoidal variation in $\Delta {T}({z})$ , which could be misinterpreted as a sinusoidal variation in the linear energy deposition rate within the OFBGC’s thermal mass. However, for the specific design of the OFBGC that was built and tested in the OSURR, for the slowly varying spatial variation of ${T}_{B}({z})$ along the length of the OSURR’s Water Irradiation Facility, within which the OFBGC was tested, the perturbation in $\Delta {T}({z})$ is negligibly small. [ABSTRACT FROM AUTHOR]

Fuel channels in CANDU® (CANadian Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). The gas annulus space between the 6 m long tubes, known as the gap, is monitored from within the PT using a transmit-receive eddy current probe that can detect changes in the environment using electromagnetic induction according to Faraday’s Law. Accurate measurements are required to ensure that contact between PT and CT is not imminent. Fuel channels may also pass over perpendicularly oriented tubes (nozzles) that are part of the liquid injection shutdown system (LISS). The proximity of LISS nozzles not only compromises gap measurement, but flow-induced vibrations from within the moderator could lead to CT deformation and/or fretting at LISS-CT contact. Investigation of the eddy current based gap probe as a tool to measure proximity of LISS nozzles was carried out experimentally. Eddy current response as a function of LISS-PT proximity, the distance between the PT outer diameter and LISS nozzle outer diameter, was determined. When PT wall thickness, PT resistivity, probe lift-off, and PT-CT gap variations spanned typical values found in-reactor this dependence could be used to determine the LISS-PT proximity up to 20 mm with sub-millimeter accuracy. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby, provide a means of optimizing maintenance schedules. [ABSTRACT FROM AUTHOR]

The Micro-Pocket Fission Detector (MPFD) is a small-form-factor real-time fission chamber. MPFD performance has been simulated in the Advanced Test Reactor Critical Facility located at Idaho National Laboratory. The neutron and gamma-ray flux profiles and magnitudes were simulated using MCNP in the near-core B-8 irradiation position. These simulations were performed at 69 discrete axial locations inside the B-8 position 55 for three separate orientations of the nearby hafnium outer shim control cylinders and at a power level of 700 W(thermal). The resulting neutron and gamma-ray flux values were used to determine the MPFD response for various fissile masses and detector gas pressures. The optimal gas-operating pressure was determined to be between 30 and 60 psig. The required fissile layer mass was determined to be between 0.5 to 1.0 µg of 235U. Additionally, the gamma ray to fission fragment interaction rate was determined to be 1.43 × 103 with average energy deposition for gamma rays and fission fragments in 30 psig argon gas to be 1 keV and 3.5 MeV, respectively. [ABSTRACT FROM AUTHOR]

Measurements and simulations of thermal neutron attenuation by printed circuit boards (PCBs) are compared. Attenuation coefficients in typical epoxy-resin/glass-fiber substrate material can be as high as 2 cm−1, corresponding to 27% attenuation by 1.6 mm of the substrate. Attenuation is attributed to neutron scattering off hydrogen in the resin acting in synergy with absorption by boron in the glass; this effect is substantially greater than that estimated from absorption by boron alone. Design of thermal neutron detector assemblies should take this attenuation into account and may require board thickness to be minimized or specialized substrate materials to be used. [ABSTRACT FROM AUTHOR]

The optical properties of FLiBe salt in a Fluoride-Salt-Cooled High-Temperature Reactor (FHR) present an opportunity to utilize Cerenkov radiation measurements to reconstruct the power profile in the core and detect various anomalies that could occur during operation. The Cerenkov light produced within a coolant channel is strongly correlated to the fission rate density and power level in the surrounding fuel assembly and travels freely through the optically transparent salt. The light coming from coolant channels can be measured by an array of photon detectors above the channels or a system of mirrors and light guides to a detector. This allows the assembly-level power profile in the core to be reconstructed, identifying hot spots within the core. By comparing the levels of light detected to a baseline operating state, anomalies can be detected as well as their location within the core. The method has been developed and assessed computationally to realize this approach for FHRs. Details of the method and demonstrations of its applications are discussed in this paper. [ABSTRACT FROM AUTHOR]

Diagnostic “Divertor Neutron Flux Monitor (DNFM)” is one of the diagnostics to determine the neutron yield and the thermonuclear power of the ITER facility. To detect neutrons, this diagnostic uses three detector modules. The expected neutron flux in the detector module location is 106–1013 n ⋅ cm−2 ⋅ s−1. The neutron flux and the neutron yield should be measured with 1-ms time resolution and 10% maximum relative error. The detector module consisting of six fission chambers (FCs) is used to meet these requirements, and each FC signal is processed by pulse counting, Campbell, and current measurement methods simultaneously. According to the data of each measurement method, the normalized count rates (NCRs) of the FCs and the detector module and the neutron flux at its location are calculated. The data acquisition (DAQ) system processes the signals of the single detector module and transmits the data of the NCR and the neutron flux of each detector and entire detector module to the upper-level system. To verify the technique solutions, the DAQ system prototype was manufactured. To study the characteristics of this prototype, a set of laboratory tests was carried out. This article illustrates the structure of the DAQ system and the implementation of the FC signal processing methods. The results of the study of the DAQ system prototype for a single detector module are given. [ABSTRACT FROM AUTHOR]