Detection Of Delayed Neutrons Research Articles

Conception of delayed neutron detector blocks in a sodium cooled fast reactor

The primary vessel of a sodium-cooled fast neutron reactor (SFR) must remain free of any contaminant released by a cladding failure in a fuel assembly. A comprehensive detection system of cladding failures has then to be considered right from the beginning of the reactor design. One sub-system, called delayed neutron detection (DND), aims to search for delayed neutron precursors (DNP) through an ex-vessel analysis of a sodium sample, which goes to a container, usually named a pot, surrounded by neutron detectors. This paper presents a methodology to optimize the hydraulic part of the design of the DND detection block. From a detection scenario that identifies the physical parameters in play, a computational model is derived, which couples fluid mechanics and neutron transport. Different types of pot designs are considered, for which the sodium flow velocity, DNP concentration and detector activities are assessed. A new type of pot with a coil shape is proposed and shown to have the better performances with respect to the detector response to delayed neutrons.

A Novel Assessment of Delayed Neutron Detector Data in CANDU Reactors

Abstract A common opportunity for nuclear power plant operators is ensuring that routinely collected data are fully leveraged. Exploiting data analytics can enable improvements in anomaly detection and condition monitoring by identifying previously unseen data trends and correlations without major financial investment. One such opportunity is in facilitating the detection of fuel defects by augmenting the delayed neutron (DN) monitoring system deployed in the majority of Canada deuterium uranium (CANDU) reactors. In this paper, we demonstrate using archive data that the detection of fuel defects can be accelerated using this system in combination with the use of a deeper historical dataset and the introduction of a smoothing algorithm. The current defect identification process relies on the analysis of data of high variance and is subject to the judgment of a domain expert, resulting in variable defect identification periods. The proposed approaches seek to mitigate this and alleviate the variable identification time. Initial results presented here show that for an initial batch of 30 defects, identification periods can be meaningfully reduced compared to the current process, with defects potentially visible on an average of 11.4 days earlier. By shortening this identification period, fuel containing defects can be scheduled for earlier removal, reducing the risk of statutory shutdown obligations, protecting personnel, and promoting industry best practice. Exploring a historical dataset identifies previously undocumented trends and we discuss the potential to produce correlations with other reactor parameters. The application of this knowledge can lead to opportunities in the use of machine learning algorithms and, ultimately, more accurate predictions.

Development and Validation of SAS4A Code and Its Application to Analyses on Severe Flow Blockage Accidents in a Sodium-Cooled Fast Reactor

Local subassembly faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage (HTIB) at the coolant inlet of an SA gives most severe consequences among a variety of LFs. Although an evaluation on the consequences of HTIB using SAS4A code was performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in an SFR by this developed SAS4A code was also performed in this study. It was clarified by the analyses considering power control system that the reactor would be safely shut down by the reactor protection system triggered by either of 116% over power or delayed neutron detector (DND) trip signals. Therefore, the conclusion in the past study that the consequences of HTIB would be much less severe than that of unprotected-loss-of-flow (ULOF) was strongly supported by this study. Furthermore, SAS4A code was newly validated using four in-pile experiments which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. Thus, the methodology of HTIB evaluation was established in this study together with the past study and is applicable to HTIB evaluations in other SFRs.

Flow induced vibration studies on PFBR control plug components

The construction of Prototype Fast Breeder Reactor (PFBR), a 500MWe liquid sodium cooled reactor, is in progress at Kalpakkam in India. Control plug (CP) is located right above the core subassemblies in the hot pool. Control plug is an important component as many of the critical reactor parameters are sensed and controlled by the components housed in the control plug assembly. In PFBR primary circuit, components are basically thin walled, slender shells with diameter to thickness ratio ranging from 100 to 650. These components are prone to flow induced vibrations. The existence of free liquid (sodium) surfaces, which is the source of sloshing phenomenon and the operation of primary sodium pump in the primary pool are other potential sources of vibration of reactor components.Control plug is a hollow cylindrical shell structure and provides passages and support for 12 absorber rod drive mechanisms (ARDM) which consists of 9 control and safety rods and 3 diverse safety rods, 210 thermo wells to measure the sodium temperature at the exit of various fuel subassemblies, three failed fuel localization modules (FFLM) and acoustic detectors. It consists of a core cover plate (CCP), which forms the bottom end, two intermediate supports plate, i.e. lower stay plate (LSP) and upper stay plate (USP) and an outer shell. The CCP is located at a distance of 1.3m from the core top. With such a gap, there will be long free hanging length of the thermocouple sleeves, Delayed neutron detector (DND) sampling tubes and ARDM shroud tubes and hence they are susceptible to flow induced vibrations. Since control plug is partially immersed in hot sodium, the reactor transients are felt by the components, hence it is very much essential to understand the vibration response of the control plug components. The main vibration excitation mechanisms are vortex shedding, turbulence buffeting, fluid-elastic instability, etc.In order to assess the susceptibility of CP internals against flow induced vibrations (FIVs), to measure structural response and to validate the analytical codes developed for FIV analysis of CP along with the assumptions of supports for various CP internals therein, a flow induced vibration testing program was formulated in 1:4 scale model of PFBR CP. As the first phase of this program, experimental modal analysis of CP internals was carried out in air to estimate the modal parameters. Subsequently, flow induced vibration studies were conducted in the 1:4 scale model of CP in water. In this model, size of the tubes, shell and plates are reduced to ¼ size of the PFBR CP. The stiffness of CP parts is reduced by 4 times and mass decreased by 64 times which results in scaling up of modal frequencies by 4 times. The CP internals in the model were instrumented with accelerometers and strain gages. The studies were conducted in water with flows derived on the basis of velocity similitude. The output signals from the sensors were acquired and analyzed to obtain frequency spectra, overall vibration amplitude and strain values at various locations inside CP. The study carried out confirms the absence of resonance due to flow induced vibration mechanisms for the entire operating range. This paper elucidates the modeling details, similitude criteria, instrumentation employed and experimental results obtained with discussion on results.

β-delayed neutron emission studies

The study of β-delayed neutron emission plays a major role in different fields such as nuclear technology, nuclear astrophysics and nuclear structure. However the quality of the existing experimental data nowadays is not sufficient for the various technical and scientific applications and new high precision measurements are necessary to improve the data bases. One key aspect to the success of these high precission measurements is the use of a very pure ion beam that ensures that only the ion of interest is produced. The combination of the IGISOL mass separator with the JYFLTRAP Penning trap is an excellent tool for this type of measurement because of the ability to deliver isobarically and even isomerically clean beams. Another key feature of the installation is the non-chemical selectivity of the IGISOL ion source which allows measurements in the important region of refractory elements. This paper summarises the β-delayed neutron emission studies that have been carried out at the IGISOL facility with two different neutron detectors based on 3He counters in a polyethylene moderator: the Mainz neutron detector and the BEta deLayEd Neutron detector.

First Measurements with the BEta deLayEd Neutron Detector (BELEN-20) at JYFLTRAP

A prototype version of the BEta deLayEd Neutron detector, which is being developed for the FAIR/DESPEC experiment, has been used for the first time in an experiment at JYFL (Finland). This detector is based on 3He counters and its first run was primarily intended to commission the detector and verify the working principles for future experiments. A new triggerless DAQ has been developed for these measurements. This DAQ time-stamps the events and allows complete flexibility to construct correlations offline.An isotopically pure beam was obtained using the JYFLTRAP Penning trap setup at the IGISOL facility and it was implanted on a movable tape placed in the centre of the BELEN-20 detector. The measurements were performed for known delayed neutron emitters of interest for nuclear power generation: 88Br, 94,95Rb, 138I.The characteristics of the BELEN-20 detector and the first experiment at JYFL are presented in this paper.

Monte Carlo Simulations for the Study of a Moderated Neutron Detector

This work presents the Monte Carlo simulations performed with the MCNPX and GEANT4 codes for the design of a BEta deLayEd Neutron detector, BELEN-20. This detector will be used for the study of beta delayed neutron emission and consists of a block of polyethylene with dimensions 90x90x80 cm3 and 20 cylindrical 3He gas counters. The results of these simulations have been validated experimentally with a 252Cf source in the laboratory at UPC, Barcelona. Also the first experiment with this detector has been carried out in November 2009 in JYFL, Finland. In this experiment the neutron emission probability after beta decay of the fission products 88Br, 94;95Rb, and 138I has been measured; this data is still under analysis. Simulations with MCNPX and GEANT4 have been performed in order to obtain the efficiency of the BELEN-20 detector for each of the above nuclei using the neutron energy distribution corresponding to each nucleus.

Benchmark validation comparisons of measured and calculated delayed neutron detector responses for a pulsed photonuclear assessment technique

An MCNPX-based calculational methodology has been developed to numerically simulate the complex electron–photon–neutron transport problem for the active interrogation system known as the pulsed photonuclear assessment (PPA) technique. The PPA technique uses a pulsed electron accelerator to generate bremsstrahlung photons in order to fission nuclear materials. Delayed neutron radiation is then detected with helium-3 neutron detectors as evidence of the nuclear material presence. Two experimental tests were designed, setup and run to generate experimental data for benchmarking purposes. The first test irradiated depleted uranium in air, and the second test, depleted uranium in a simulated cargo container (plywood pallet), using 10MeV electron pulses. Time-integrated, post-flash, delayed neutron counts were measured and compared to calculated count predictions in order to benchmark the calculational methodology and computer models. Comparisons between the experimental measurements and numerical predictions of the delayed neutron detector responses resulted in reasonable experiment/calculated ratios of 1.42 and 1.06 for the two tests. High-enriched uranium (HEU) predictions were also made with the benchmarked models.

Spectroscopy around N = 20 shell closure: β–n decay study of 35Al

An experimental spectroscopic investigation of 35Al has been performed. The delayed neutron and γ spectra, following the β decay of 35Al, have been measured using the delayed neutron detector array TONNERRE coupled with high efficiency γ detectors. A detailed analysis of such spectra allows one to extract the level schemes of its daughters. These levels are compared with the few previously existing data and shell-model calculations. Based on triple β–γ–n coincidences eight new levels above the neutron separation energy in 34Si were for the first time observed.

Derivation of Nuclear Parameters for Delayed Neutron Detector Measurements for D-D and D-T Plasma Operation at the Joint European Torus

The first attempt to calculate the parameters for 238U and 232Th used in the analysis of delayed neutron counter measurements of the total neutron yield from deuterium-tritium (D-T) plasmas is desc...

Sodium In-Pile Loop Experiment on Delayed Neutron Detector-, Precipitator-, and Cover-Gas Gamma-Ray Spectrometer-Type Fuel Failure Detection Systems

Performance of four different types of fuel failure detection systems has been studied using a sodium in-pile loop located in Japan Research Reactor No. 2. In the sodium in-pile loop, 8.5 liter sodium circulates through an irradiation section (500°C), a main cooler (400°C), an expansion tank (380°C), an electromagnetic pump, and a main heater at a flow rate of 3 liters/min. In the irradiation section, four metallic uranium (20% enriched) plates of a total amount of U-235 of 9.18g had been irradiated at a thermal neutron flux of 5.4×1010 n/cm2sec. A delayed neutron detector consisting of a BF3 counter system in a graphite moderator (50cm×5Ocm×8Ocm) detects delayed neutrons from fission products in the expansion tank. Helium cover gas over the sodium in the expansion tank carries gaseous fission products to a precipitator (The Plessey Co. Ltd., Mark XI) by which beta-rays from Rb-88 and Cs-138 collected on the precipitator wire are detected. The precipitator was modified to measure gamma-ray spectra from the wire using a Ge(Li) detector system for the nuclide identification. The helium gas is also guided into a gas reservoir whose gamma-ray spectra are analysed by another Ge(Li) detector system. Construction and performance of these detection systems was described in detail and their detection sensitivities and response were compared.