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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.

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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.

Small fission chambers manufactured by Photonis, Reuter-Stokes, CEA, and Centronic are routinely calibrated in standard neutron fields of the BR1 reactor at SCK CEN in Mol, Belgium. Two irradiation fields are available: fast spectrum in the MARK3 convertor and thermal spectrum in the Empty Cavity. In this work, for the first time the calibration procedure of larger fission chambers (Photonis CFUL01 and CFUM21 type) with the deposit length exceeding the length of the convertor is presented. Spatial energy correction factors taking into account the neutron flux shape beyond the convertor and radial gradient of the neutron flux were calculated with MCNP and experimentally validated. The combination of calibration (i.e., effective mass measurement) in fast and thermal irradiation fields allows for determination of the purity of 238U deposit. This is demonstrated on three CFUL01-type fission chambers with purity between 99.8% and 99.998%. [ABSTRACT FROM AUTHOR]

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.

This article presents a procedure for optimizing the charge comparison method (CCM) used for pulse shape discrimination (PSD). Without prior knowledge of the signals or the readout system, our procedure automatically optimizes the integration periods maximizing the discrimination ability of the radiation detector. This procedure is innovative in its adaptability and automation without being complicated to implement on a standard computer. Another advantage of this approach is the possibility to use it even if the operation of the readout system and the recording process of the signal is not fully known. Therefore, it enables all detection systems generating signals whose temporal evolution depends on the origin to optimize the integration periods of the CCM. Our procedure is based on verifying that two criteria are met in terms of the number of components and the correlation of Gaussian fits made on the distribution of the tail-to-total integral resulting from the CCM. We tested the procedure for different application cases. First, the optimization of the integration periods of the CCM was performed for the discrimination between fast neutrons and gamma rays with a plastic scintillator and a silicon photomultiplier (SiPM) in the energy range [250 keVee; 4.5 MeVee]. The integration periods, from the laboratory’s experience with photomultiplier tubes (PMTs) and plastic scintillators, gave a Figure of Merit (FoM) of 0.58 corresponding to a rejection ratio (RR) of 8.6%. The procedure improved the FoM up to 0.88 corresponding to a RR of 1.9%. We also applied the procedure to the discrimination between beta and gamma rays with a PMT and a phoswich organic detector and to the discrimination between signals collected from neutrons or partial discharges within a fission chamber. [ABSTRACT FROM AUTHOR]

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.

• This paper proposed a novel automatic calibration method which can automatically deduct various baseline of Campbelling mode and obtain neutron yield in real time. • The automatic calibration method was verified with the experimental data on EAST. It is suitable for different experiment conditions, and the Relative Root Mean Square Error values of automatic calibration results were obviously lower than that of offline fixed calibration results. To satisfy the wide-range requirement for the fission chamber-based neutron flux monitoring (NFM) system on experimental advanced superconducting tokamak (EAST), the pulse-counting and Campbell modes were combined to expand the measurement range. However, to obtain the neutron yields, the results of the Campbell mode should first be converted into neutron count rates via linear calibration. During the plasma experiment, the non-neutron backgrounds vary with respect to different experimental conditions, which leads to a baseline dynamic change in the Campbell mode. In future research, NFM system need provide real-time feedback parameters regarding the operational status of the fusion device. Therefore, an automatic calibration method was developed to automatically deduct various baselines of the Campbell mode and obtain the neutron yields in real time. This method continuously updates the average value of each sub-interval (multiple sub-intervals are set in the calibration region) using a linear superposition averaging method. In addition, to optimize the dual-mode switching threshold, ensure the calibration region and define the upper limit of Campbell mode, a simulated neutron signal generator was designed and applied. The switching threshold of the dual modes was set to 400 kcps, the calibration region of the Campbell mode was selected as 50–500 kcps, and the upper limit of the equivalent count rate in the Campbell mode reached to 10 8 cps with a relative error of<4%. Finally, the automatic calibration method was tested using experimental data from the NFM system on EAST. The automatic calibration results basically overlapped with the reference results and superior to the offline fixed calibration results. Under different experimental conditions, the relative root mean square error (RRMSE) values of the automatic calibration results were lower than those of the offline fixed calibration results. Therefore, the proposed automatic calibration method is accurate and reliable, and can satisfy the real-time measurement requirements of wide-range NFM system. [ABSTRACT FROM AUTHOR]

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.

In order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its research and development effort in the field of in-pile instrumentation during the last decade. Major progress has been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. A significant part of this effort has been conducted in the framework of the Joint Instrumentation Laboratory between the CEA and the Belgian Nuclear Research Centre (SCK $\cdot$ CEN). 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. The fast-neutron-detection-system (FNDS) has been designed to monitor accurately high-energy neutron flux (E > 1 MeV) in typical material testing reactor (MTR) conditions, where neutron flux levels can be as high as $10^{15}\,\,\text {N}\cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ and are generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few $10^{15} \gamma \cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ , which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors allowing the simultaneous detection of both thermal- and fast-neutron flux. Thermal neutrons can be measured using a self-powered neutron detector or a 235U miniature fission chamber, while fast neutron detection requires 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 $\cdot$ CEN in Belgium. Two FNDS prototypes were operated in-pile during nearly 1000 h. These tests exhibited the consistency of the measurement of thermal to fast neutron flux ratio with Monte Carlo neutron and photon transport calculations, as well as the right compensation of fissile material depletion. This paper describes a second test recently completed at ISIS reactor operated by CEA in France. For this irradiation, FNDS signal was compared to reference thermal- and fast-neutron flux measurements using activation dosimeters analyzed under COFRAC Quality Certification. During this latter test, FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%. FNDS is now operational and is assumed to be the first and unique acquisition system able to provide an online measurement of the fast neutron flux in MTR conditions. This system will be used to perform spectral neutron characterization of JHR channels, but it may also be implemented in future irradiation experiments, for a better and real-time evaluation of the fast neutron flux received by material and fuel samples. [ABSTRACT FROM AUTHOR]

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.

In this study, we deal with the neutron flux monitoring inside the TRIGA MARK II reactor as one of the nonnegative matrix factorization problems. The fact that these methods of separating blind sources does not require any assumptions on the way the signal and noise were mixed has encouraged us to apply them to online neutron flux monitoring through neutron-gamma discrimination. We synthesize fission chambers output signals using the Python-based Fission Chamber simulator (pyFC). These signals are then considered as time series mixtures that are analyzed through nonnegative algorithms known as the most popular blind source separation techniques to extract original sources signals. The obtained results show that the Regularized Lee–Sung algorithm is the most appropriate method for discriminating neutrons and gammas. Furthermore, the neutron-gamma discrimination is highlighted through the computation of the cross-correlation functions between the separated source signals and pure neutron and gamma signals. The presented method provides a better characterization of the neutron and gamma signals with very high accuracy. [ABSTRACT FROM AUTHOR]

Within the bilateral project between the CEA Cadarache and the Jožef Stefan Institute (JSI) a wide variety of measurements using multiple fission chambers simultaneously inside the reactor core were performed. The fission rate axial profiles were measured at different positions in the reactor core and at different control rod configurations. A relative comparison of the calculated fission rates using the MCNP code and the measured fission rates was performed. In general the agreement between the measurements and calculations is good, with the deviations within the uncertainties. For better observation and understanding of the neutron flux redistribution due to the control rod movement, the neutron flux and fission rate had been calculated through the entire reactor core for different control rod configurations. The detector position with minimum signal variations due to the regulating and compensating control rod movement during normal operation was determined. The minimum variation is optimal in case we want to reliably determine the reactor power without influence of the regulating and compensating control rod positions. [ABSTRACT FROM AUTHOR]

The Large Helical Device (LHD) plans to start deuterium experiments in March 2017. The expected performances of the neutron flux monitor and the vertical neutron camera have been investigated by neutron Monte Carlo simulations using the MCNP-6 Monte-Carlo neutronics code and the cross-section library FENDL-3.0. Three neutron flux monitors using 235 U fission chambers are located on the center axis of LHD and at two toroidal locations on the horizontal plane outside the cryostat. The detection efficiency for the neutron emission in the plasma has been evaluated. This is almost insensitive to the neutron source profile and the plasma position. The vertical neutron camera consists of 11 channels of stilbene detectors with a parallel collimator embedded in the concrete floor slab. The response function of the stilbene detector has been evaluated with PHITS code. Based on the response function, the count rate of the stilbene detector is estimated to be ∼6 × 10 4 cps for maximum neutron yield shots. Also, the spatial resolution is estimated to be 70 mm, which is smaller than the pitch of adjoining channels. [ABSTRACT FROM AUTHOR]

• Systematic overestimation discovered in previously reported fission chamber neutron sensitivity data. • Tallying neutron flux from Maxwellian distribution source in MCNP was identified as the source of error. • New procedure to properly simulate neutron moderation was implemented. • Corrected results are reported. During the course of the work carried out for the estimation of the sensitivity of 6Li loaded fiber detectors (Borella, 2022), it was realized that the MCNP simulations carried out with the approach described in (Borella et al., 2013) were affected by a systematic bias. In this paper, we explain the source of the systematic bias and report the new results for (Borella et al., 2013). [ABSTRACT FROM AUTHOR]

• The moderator of the NFM#01 system has been determined according to neutronic analysis results. • NFM system is capable of measuring a wide range of total neutron yield and fusion power. • Measurement performance of NFM system can meet ITER requirements well based on current physical design. The ITER Neutron Flux Monitor (NFM) system consists of four subsystems located at ITER Equatorial Ports 1#, 7#, 8#, and 17#. NFM system carries out the measurement of total neutron yield and fusion power by employing 235U fission chambers. The total neutron yield to be measured spans a range of 7 orders of magnitude — from 1014 to 3.2 × 1020 n/s. Suitable moderator and detectors were selected to cover such a wide range of neutron yield and meet the requirements of time resolution and accuracy. Moderators and detectors' sensitivities of NFM system are reasonably designed. The detectors' sensitivities of the NFM#01 system under different moderators have been calculated with MCNP code, and graphite serves as the optimal moderator. The detectors' sensitivities of the other three NFM subsystems have also been simulated simultaneously. Based on the required sensitivities of detector, the appropriate mass of 235U was selected for each fission chamber. According to the analysis of the measurement range of each NFM subsystem, it is finally confirmed that the measurement performance of the NFM system can meet the requirements well. [ABSTRACT FROM AUTHOR]

Via thermal cycling and vibration tests, this study aims to demonstrate that the in-vessel components of the microfission chamber (MFC) system can withstand the extreme International Thermonuclear Experimental Reactor (ITER) environment. In thermal cycle tests, the signal cable of the device was bent into a smaller radius and it was given more bends than those in its actual configuration within ITER. A faster rate of temperature change than that under the typical ITER baking scenario was then imposed on in-vessel components. For the vibration tests, strong 10 G vibrational accelerations with frequencies ranging from 30 Hz to 2000 Hz were imposed to the detector and the connector of the in-vessel components to simulate various types of electromagnetic events. Soundness verification tests of the in-vessel components conducted after thermal cycling and vibration testing indicated that problems related to the signal transmission cable functioning were not found. Thus, it was demonstrated that the in-vessel components of the MFC can withstand the extreme environment within ITER. [ABSTRACT FROM AUTHOR]

Applying Higher Order Campbelling methods in neutron flux monitoring with fission chambers is advantageous due to their capabilities to suppress the impact of unwanted noises and signal contributions (such as gamma radiation). This work aims to verify through experimental results that the basic assumptions behind the Higher Order Campelling methods are valid in critical reactors. The experiments, reported in this work, were performed at the MINERVE reactor in Cadarache. It is shown that the calibration of a fission chamber and the associated electronic system is possible in higher order mode. With the use of unbiased cumulant estimators and with digital processing, it is shown that over a wide count rate range, accurate count rate estimation can be achieved based on signal samples of a few ms , which is a significant progress compared to similar experimental results in the literature. The difference between the count rate estimated by pulse counting and by the Higher Order Campelling is less than 4%. The work also investigates the possibility of monitoring transient events. For this purpose, a control rod drop event was followed in Higher Order Campbelling mode. [ABSTRACT FROM AUTHOR]

New fission chambers dedicated to Prompt Fission Neutron Spectra measurements with the time-of-flight technique have been developed. The actinide mass embedded in the chamber was maximized, while the alpha-fission discrimination and the time resolution were optimized. Moreover, to reduce the neutron background and spectra distortions, neutron scattering with the materials were minimized by the choice of material and structure. These chambers were then tested and validated during tests and in-beam experiments. [ABSTRACT FROM AUTHOR]

Fission chambers have become one of the main devices for the measurement of neutron fluxes in nuclear facilities; including fission reactors, future fusion ones, spallation sources, etc. The main goal of a fission chamber is to estimate the neutron flux inside the facility, as well as instantaneous changes in the irradiation conditions. A Monte Carlo Fission Chamber Designer (MCFCD) has been developed in order to assist engineers in the complete design cycle of the fission chambers. So far MCFCD focuses on the most important neutron reactions taking place in a thermal nuclear reactor. A theoretical model describing the most important outcomes in fission chambers design has been developed, including the expected electrical signals (current intensity and drop in potential) and, current-polarization voltage characteristics (sensitivity and saturation plateau); the saturation plateau is the zone of the saturation curve where the output current is proportional to fission rate; fission chambers work in this region. Data provided by MCFCD are in good agreement with measurements available. [ABSTRACT FROM AUTHOR]

This paper investigates, through numerical simulations, the performance of a signal analysis method by which a high temperature fission chamber can be used over a wide range of count rates. Results reported in a previous paper (Elter et al., 2015 [1] ) indicated that the traditional Campbell method and the pulse mode cannot provide a sufficient overlap at medium count rates. Hence the use of the so-called Higher Order Campbell (HOC) methods is proposed and their performance is investigated. It is shown that the HOC methods can guarantee the linearity (i.e. correctness) of the neutron flux estimation over a wide count rate, even during transient conditions. The capabilities of these methods for suppressing parasitic noise (originating from various sources) are verified. [ABSTRACT FROM AUTHOR]

Preparation of thin U- and Th-coated 0.3 mm diameter Pt working electrodes by the cyclic potential sweep method is described. Uranyl- and thorium hydroxide layers were electrodeposited from ethanol solutions containing 0.02 M natural uranyl and 0.02 M natural thorium nitrate, each with 3.6 M ammonium nitrate. The cell for electrodeposition was specially developed in order to accommodate the small working electrodes for this research by including a working electrode probe, 3-D translation stage, and microscope. The source material deposition was analyzed using digital microscopy and scanning electron microscopy, and confirmed using x-ray fluorescence measurements. The appropriate potential range for electrodeposition was determined to be −0.62 V to −0.64 V for a 0.3 mm diameter Pt working electrode placed 1 cm from the counter electrode. Smooth, uniform deposition was observed near the central region of the working electrode, while surface cracking and crystalline formations were found near the edge of the working electrode. The final procedure for sample substrate preparation, electrolytic solution preparation and electrodeposition are described. [ABSTRACT FROM AUTHOR]

Neutron sensors capable of real-time measurement of neutrons in high-flux environments are necessary for tests aimed at demonstrating the performance of experimental nuclear reactor fuels and materials in material test reactors (MTRs). In-core Micro-Pocket Fission Detectors (MPFDs) have been studied at Kansas State University for many years. Previous MPFD prototypes were successfully built and tested with promising results. Efforts are now underway to develop advanced MPFDs with radiation-resistant, high-temperature materials capable of withstanding irradiation test conditions in high performance material and test reactors. Stackable MPFDs have been designed, built, and successfully demonstrated as in-core neutron sensors. Advances in the electrodeposition and measurement of neutron reactive material, along with refinements to composition optimization simulations, have enhanced the capabilities of contemporary MPFDs. [ABSTRACT FROM AUTHOR]

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. [ABSTRACT FROM AUTHOR]

This work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E \sim 0.97). [ABSTRACT FROM PUBLISHER]

The detectors of the neutron flux monitoring system of the foreseen French GEN-IV sodium-cooled fast reactor (SFR) will be high temperature fission chambers placed in the reactor vessel in the vicinity of the core. The operation of a fission chamber over a wide-range neutron flux will be feasible provided that the overlap of the applicability of its pulse and Campbelling operational modes is ensured. This paper addresses the question of the linearity of these two modes and it also presents our recent efforts to develop a specific code for the simulation of fission chamber pulse trains. Our developed simulation code is described and its overall verification is shown. An extensive quantitative investigation was performed to explore the applicability limits of these two standard modes. It was found that for short pulses the overlap between the pulse and Campbelling modes can be guaranteed if the standard deviation of the background noise is not higher than 5% of the pulse amplitude. It was also shown that the Campbelling mode is sensitive to parasitic noise, while the performance of the pulse mode is affected by the stochastic amplitude distributions. [ABSTRACT FROM AUTHOR]

CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. [ABSTRACT FROM AUTHOR]