Search

In order to overcome the difficulties of meeting the engineering application requirements of the technical performance indicators of the fission ionization chamber, to explore the feasibility of the development of the longsensitive zone fission ionization chamber in actual application, and to realize the domestic application of the intermediate range detector for the excore nuclear instrumentation system (NIS) in the third-generation nuclear power plant, a cylindrical fission ionization chamber for the nuclear power plant external nuclear measurement system was developed. Based on the thermal neutron fluence standard device, standard γ radiation device and reactor irradiation hole test platform, the identification characteristics, thermal neutron sensitivity, charge per pulse, average electron collection time, high voltage plateau characteristics, repeatability and γ sensitivity of the fission ionization chamber were tested, and the relationship between the detector performance and temperature was tested. The test results show that the thermal neutron sensitivity of the fission ionization chamber can reach 0.53 s-1/nv, the charge per pulse is 1.88×10-13 C, the average electron collection time is 400 ns, the high voltage platform characteristics platform length ≥400 V, platform slope ≤4%/100 V, γ sensitivity 3.30×10-9 A·Gy-1·h, repeatability maximum value 2.4%, all indicators are better than the design requirements. The successful development of the fission ionization chamber lays a solid foundation for the subsequent development of the long-sensitive zone fission ionization chamber.

KHNP had licensed Dynamic Control rod Reactivity Measurement (DCRM) method using detector current signals of PWRs in 2006. The method has been applied to all PWRs in Korea for about 15 years successfully. However, the original method was inapplicable to PWRs using low-sensitivity integral fission chamber as ex-core detectors because of their pulse pile-up and the nonlinearity of the mean-square voltage at low power region. Therefore, to overcome this disadvantage, a modified method, DCRM-EK, was developed using kinetics behavior after equilibrium condition where the pulse counts maintain the maximum value before pulse pile-up. Overall measurement, analysis procedure, and related computer codes were changed slightly to reflect the site test condition. The new method was applied to a total of 15 control rods of 1000 MWe and 1400 MWe PWRs in Korea with worths in the range of 200 pcm–1200 pcm. The results show the average difference of -0.4% and the maximum difference of 7.1% compared to the design values. Therefore, the new DCRM-EK will be applied to PWRs using low sensitivity integral fission chambers, and also can replace the original DCRM when the evaluation fails by big noises present in current or voltage signals of uncompensated/compensated ion chambers.

The CEA Sensor, Dosimetry and Instrumentation Laboratory (LDCI) has been working since 2011 on an integrated acquisition system called MONACO, ’Multichannel Online Neutron Acquisition in Campbell mOde’, for fission chamber measurements. Taking into account the feedback acquired up to the CEA TRL7 version, the LDCI Lab and the Instrumentation Technologies company (Solkan – Slovenia) initiated in 2020 a two years collaboration agreement to build the Libera MONACO 3 industrialized version required for reactor use and mass production. After a final qualification performed in the Slovenian TRIGA Mark II reactor in March 2023, the system will be available on the nuclear instrumentation market by the end of 2023.

Real-time characterization of irradiation facilities improves the utilization of the core capabilities of test nuclear reactors. The ability to observe how the local neutron flux (level and spectrum) changes as control elements and experiments change will fundamentally transform our understanding of the underlying physical phenomena that govern the operation of present and advanced nuclear reactors, ultimately providing valuable information for the nuclear energy industry. The objective of this research was to demonstrate how advanced sensors could be used to significantly reduce the time and cost of experiments, improve our understanding of experimental environments, and enable verification and validation of simulation and modeling methods. This was accomplished by designing and fabricating a dedicated real-time instrument test train for the Advanced Test Reactor Critical (ATR-C) facility. The first year of this project focused on the design and modeling of real-time axial neutron flux monitors, leveraging proven technologies pioneered at the Idaho National Laboratory, to characterize the transient that occurs in the Small-B positions at the Advanced Test Reactor and the Advanced Test Reactor Critical Facility. We found that the flux amplitude in those positions can fluctuate as much as 380% depending on the outer shim control cylinder position. The engineering design of the test fixture and flux monitor instrumentation was the objective of the 2nd project year. New capabilities were established to electrodeposit enriched uranium for fission chamber development at the Idaho National Laboratory and trials were begun to characterize the process. The final year included the fabrication of the test fixture and instruments for Advanced Test Reactor Critical Facility. The fabrication process was delayed by supply chain and personal availability caused by the COVID-19 pandemic. However, we were still able to deliver this unique capability to Advanced Test Reactor Critical Facility that will enable future instrument testing and scientific experiments.

As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.

The Jordan Research and Training Reactor (JRTR) is the first research reactor in Jordan, the commissioning of which is ongoing. The reactor is a 5-MWth, open-pool type, light-water-moderated, and cooled reactor with a heavy water reflector system. The neutron measurement system (NMS) applied to the JRTR employs a wide-range fission chamber that can cover from source range to power range. A high-sensitivity boron trifluoride counter was added to obtain more accurate measurements of the neutron signals and to calibrate the log power signals; the NMS has a major role in the entire commissioning stage. However, few case studies exist concerning the application of the NMS to a research reactor. This study introduces the features of the NMS and the boron trifluoride counter in the JRTR and shares valuable experiences from lessons learned from the system installation to its early commissioning. In particular, the background noise relative to the signal-to-noise ratio and the NMS signal interlock are elaborated. The results of the count rates with the neutron source and the effects of the discriminator threshold are summarized.

Constant improvements of the computational power and methods as well as demands of accurate and reliable measurements for reactor operation and safety require a continuous upgrade of the instrumentation. In particular, nuclear sensors used in nuclear fission reactors (research or power reactors) or in nuclear fusion facilities are operated under intense mixed neutron and gamma-ray fields, and need to be calibrated and modeled to provide selective and accurate neutron and gamma-ray measurements. The French Atomic Energy and Alternative Energies Commission (CEA) and the Jožef Stefan Institute (JSI) have started an experimental program dedicated to a detailed experimental benchmark with analysis using Monte Carlo particle transport calculations and a series of neutron and gamma-ray sensor types used in the JSI TRIGA Mark II reactor. CEA has setup a simplified TRIPOLI-4® modeling scheme of the JSI TRIGA reactor based on the information available in the IRPhEP benchmark in order to facilitate analysis of future neutron and gamma-ray measurements. These allow the CEA to perform a TRIPOLI-4 instrumentation calculation scheme benchmarked with the JSI MCNP model. This paper presents the main results of this CEA calculation scheme application and the analysis of their comparison to the JSI results obtained in 2012 with the MCNP5 & ENDF/B-VII.0 calculation scheme. This paper will conclude with some information about the new experimental program to be carried out in 2022 in the TRIGA reactor core.

The availability of neutron fields with a high neutron flux, suitable for irradiation testing of nuclear instrumentation detectors relevant for applications in nuclear facilities such as material testing reactors (MTRs), nuclear power reactors and future fusion reactors is becoming increasingly limited. Over the last several years there has been increased interest in the experimental capabilities of the 250 kW Jožef Stefan Institute (JSI) TRIGA research reactor for such applications, however, the maximal achievable neutron flux in steady-state operation mode falls short of MTR-relevant conditions. The JSI TRIGA reactor can also operate in pulse mode, with a maximal achievable peak power of approximately 1 GW, for a duration of a few ms. A collaboration project between the JSI and the French Atomic and Alternative Energy Commission (CEA) was initiated to investigate absolute neutron flux measurements at very high neutron flux levels in reactor pulse operation. Such measurements will be made possible by special CEA-developed miniature fission chambers and modern data acquisition systems, supported by the JSI TRIGA instrumentation and activation dosimetry. Additionally, measurements of the intensity of Cherenkov light are proposed and being investigated as an alternative experimental method. This paper presents the preparatory activities for an exhaustive experimental campaign, which were carried out in 2019-2020, consisting of test measurements with not fully appropriate fission chambers, activation dosimetry and silicon photomultipliers (SiPMs) The presented results provide useful and promising experimental indications relevant for the design of the experimental campaign.

The paper presents the integrated neutron spectrum analysis as a potential method for estimating the contents of fissile isotopes in SFAs. Two method implementation variants are described: (1) measurement of SFA average transmission and (2) measurement of sample average transmission in the spectrum that has passed a SFA. The authors describe the dependences of SFA average transmission on its content of the required isotope obtained by means of two types of detectors: helium counter tube and fission chamber. Also, the authors propose a method to estimate SFA burn-up by means of the integrated NSA. In addition, SFA residence time influence on transmission is estimated.

The CEA MONACO v2 a multichannel acquisition system dedicated to neutron and gamma measurements. It is unique as it integrates all the following features in one module: automatic generation of saturation curves, automatic generation of pulse discrimination curves, detector pulse characterization using the embedded oscilloscope module, pulse mode acquisitions in count rate or pulse height analysis, fluctuation mode and current mode acquisitions. Sensors are plugged to a single connector and the implemented operating modes run constantly in parallel. Firsly designed for on line local neutron and gamma measurements with gaseous detectors in reactor experiments, the MONACO v2 system will also be available for self-powered detectors thanks to its wide current mode working range. After two years of development, CEA teams tested two MONACO v2 prototypes in the Slovenian TRIGA Mark II research reactor in 2018, using CEA miniature ion chambers and SPNDs. The system is now ready for industrialization to be available on the nuclear instrumentation market.

A well characterized radiation field inside a research nuclear reactor irradiation facilities enables precise qualification of radiation effects to the irradiated samples such as nuclear heating or changes in their electrical or material properties. To support the increased utilization of the JSI TRIGA reactor irradiation facilities in the past few years mainly on account of testing novel detector designs, electronic components and material samples, we are working on increasing the neutron and gamma field characterization accuracy using various modeling and measurement techniques. In this paper we present the dose field measurements using thermo-luminescent detectors (TLD’s) with different sensitivities neutron and gamma sensitivities, along with multiple ionization and fission chamber. Experiment was performed in several steps from reactor start-up, steady operation and a rapid shutdown, during which the ionization and fission chamber signals were acquires continuously, while the TLD’s were being irradiated at different stages during reactor operation and after shutdown, to also capture response to delayed neutron and gamma field. The results presented in this paper serve for validation of JSI designed JSIR2S code for delayed radiation field determination, initial results of its application on the JSI TRIGA TLD measurements will also be presented.

VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).

In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

During the operation of the Superphenix and Phenix reactors, an aberrant electrical signal was detected from the fission chambers used for neutron flux monitoring. This signal, thought to be due to partial electrical discharge (PD) is similar to the signal resulting from neutron interactions, and is generated in fission chambers at temperatures above 400 °C. This paper reports work on the characterization and localization of the source of this electrical signal in a High Temperature Fission Chamber (HTFC). The relation between the shape of the PD signal and various parameters (nature and pressure of the chamber filling gas, electrode gap distance, and fission chamber geometry) are first described. Next, experiments designed to identify the location within the chambers where the PD are being generated are presented. After verification and refinement of the results of these localization studies, it should be possible to propose changes to the fission chamber in order to reduce or eliminate the PD signal.