Reactor Power Monitoring Research Articles

Prompt-critical power excursion measurements with optical fibers in the CABRI experimental reactor

This paper presents our experimental work to assess the capability of estimating the transient-induced power distribution evolution in the CABRI experimental reactor, using Cherenkov radiation in optical fibers. The CABRI reactor is designed to produce power transients from 100 kW up to 21 GW in less than 100 ms, in order to irradiate a test fuel pin under conditions representative of a Reactivity Insertion Accident in pressurized water reactors. Because of the wide response range and short response time required to follow the reactor power evolution during a complete transient, the usual means of detection, such as ionization or fission chamber, are rendered inoperable, especially for in-core or core vicinity measurements. For this reason, we suggest measuring the Cherenkov light produced within the optical fibers. The Cherenkov light emission is linked to local electron production, which is proportional to local gamma flux through the Compton or pair production cross-section. In other words, the intensity of Cherenkov radiation is related to the photon flux intensity. Knowledge of the fission photons emitted by the reactor gives direct insight into the fission rate, hence a spatial power density distribution could be reconstructed by using the measurement of Cherenkov light at different points in the reactor. The radio-luminescence measurements taken using photodiodes show very good linearity with the reactor power monitoring system despite the transient radiation induced attenuation observed in the optical fibers. The radio-luminescent light shows steady values with around 5% of variations over the 15 transients for low OH fibers and high OH fibers. Low OH tends to overestimate the FWHM of the power peak by 1–4%.

Evaluation of uncertainty in antineutrino spectra normalization calculations for advanced nuclear reactor monitoring

Antineutrino detection systems have been envisioned as an important aspect of safeguarding the next generation of nuclear reactors, especially considering designs utilizing exotic fuel cycles. Deployment of antineutrino detection systems for safeguarding applications is hindered by the uncertainties associated with the calculations required for antineutrino spectra measurements and predictions. The focus of this work is to assess the impact of system components on antineutrino spectra normalization uncertainties and their significance in reactor power monitoring sensitivity. The dominant limitation in antineutrino detection calculations is typically the uncertainty associated with a cosmogenic background. This limitation becomes more pronounced when signals are weak, although the issue is mitigated in larger reactors due to their stronger source strength. Additionally, antineutrino emission uncertainties vary with the isotopic composition of the reactor fuel. Unconventional fuel cycles, featuring less common fissioning isotopes, such as Pu-240, introduce larger antineutrino yield uncertainties. The findings from this study suggest that future research on safeguard-targeted antineutrino detection should prioritize background mitigation, particularly when background simulation is necessary. Advanced nuclear reactor designs have a major influence on background understanding and successful system implementation.

In-Core Power Measurement Using SiC Semiconductor Detector

A SiC detector was fabricated and tested in harsh radiation environment of the High Flux Advanced Neutron Application Reactor (HANARO) reactor core at Korea Atomic Energy Research Institute (KAERI). A 4H-SiC with 30 µm thick epitaxial layer was used as the radiation sensor, and the detector was designed to be tolerable against thermal and radiation damages. Alpha response and I-V characteristics of fabricated SiC sensor was measured and detection performance of the SiC detector was evaluated in a neutron field of HANARO ex-core neutron irradiation facility. After preliminary tests, reactor power monitoring using the SiC detector was carried out by inserting it into the HANARO irradiation hole, IP-4. Radiation-induced current of the detector was recorded as reactor power increased up to 10 MWth. Maximum thermal and fast neutron fluxes were 9.4×1012 and 2.5×109 neutrons/cm2/sec, respectively, and total neutron fluence irradiated on the detector was 4.7×1016 neutrons/cm2. The detector showed good linearity of response up to the tested fluence, with R2 = 0.9997. Response speed of SiC detector was compared to that of a Rh self-powered neutron detector (SPND) in terms of signal saturation time. Averaged SiC detector saturation time was 12.8 seconds, approximately 11 times faster that of the Rh SPND.

Toward the implementation of self-powered, wireless, real-time reactor power sensing

The successful development of real-time, self-powered, wireless reactor power sensing technology would enable key operating parameters of every fuel assembly in a commercial reactor core to be continuously monitored, and potentially be utilized for applications such as closed-loop reactor control. This paper discusses the expansion of successful work on the application of Self Powered Neutron Detectors (SPND) for reactor power monitoring and the wireless transmission of these detector signals using advanced vacuum micro-electronic (VME) transmitter technology developed by Westinghouse in collaboration with the Penn State University Radiation Science and Engineering Center. Results demonstrate that this technology may be successfully applied via long-term testing in the Penn State TRIGA reactor. SPND are inherently slower than, for example, ion chambers due to their dependence on radioactive decay to generate a signal. Additional research developed delay compensation algorithms to accelerate the response of these detectors for use in near-real-time closed-loop reactor control. Initial results indicate that practical signal acceleration can be obtained digitally or with an equivalent analog circuit. Future research will apply an inverse model of the sensor’s balance differential equations to more-effectively accelerate the SPND signal. Use of inverse models will also prove to be adaptive in the sense that time-dependent burn-up of these detectors, due to neutron activation of the detector material, can be modeled and applied in the acceleration algorithm.

Basic consideration of a nuclear power monitoring system using neutron-induced prompt gamma rays

ABSTRACTMonitoring only neutron flux in a nuclear reactor core has an advantage of offering reactor power monitoring accuracy. We started the development of a new nuclear instrumentation based on the measurement of prompt gamma rays emitted from metals placed at the neutron flux monitoring positions. The thermal neutron flux at the position of each placed metal piece can be monitored by measuring the prompt gamma rays as the count rate of each energy. The gamma-ray energy range was limited from 6 to 10 MeV to mitigate the interference of environmental gamma rays. Four metals, Ti, V, Ni, and Cu, were chosen as candidates in consideration of their neutron emission rates and self-absorption. In an experiment with a high-purity germanium semiconductor detector, we considered the identification of individual peak energies in an assumed situation where prompt gamma rays were emitted from the four different metals at the same time. Energy resolutions of the peak with the largest energy gap from the nearest energy peak of the other candidate metals were smaller than the gap. Thus, we confirmed that at least one peak for each candidate metal was able to be separated from the peaks derived from other candidate metals.

Study of the PWR ex-core detector response simulation based on the 3D SN method

Simulation of the ex-core detector response function (DRF) is of great importance for reactor power monitoring, control and protection systems. However, the 3D DRF calculation requires vast majorities of computational resources due to its characteristics such as deep penetration, strong anisotropic scattering, large computational geometry, etc. In this paper, the step characteristics (SC) spatial differencing scheme and GMRES algorithm have been implemented in the 3D SN code to ensure the simulation efficient. The adjoint first-collision (FC) source method is developed to mitigate the ray effect in the DRF simulation. The Kobayashi benchmarks are calculated to verify the correctness of the FC module. Numerical results based on the Kori-1 reactor ex-core detector indicate that the SN-FC method has the comparable accuracy with multi-group Monte-Carlo method but achieves much higher efficiency.

Reactor power monitoring using Cherenkov radiation transmitted through a small-bore metallic tube

The use of a small-bore highly reflective metallic tube has been demonstrated at a pool-type nuclear reactor for transmitting Cherenkov radiation emitted from a quartz cylinder placed near the reactor core. The study revealed the promising prospect of using a metallic tube to remotely make localized measurements in order to independently monitor reactor power using Cherenkov light generated in the water pool or inside an attached radiator. This study confirmed qualitatively and quantitatively that commercially available stainless steel tubes with highly reflective inner surfaces can adequately transport Cherenkov light, produced in the proximity of a SLOWPOKE-2 reactor core operated at 20 kW, with acceptable transmission losses over significant distances. It was demonstrated that this Cherenkov-based instrument has a fast response, high fidelity, and the Cherenkov-induced output current varies linearly with the reactor power. Scans in the orthogonal directions showed the possibility of using the Cherenkov-based instrument for 3-D mapping of the general gamma field around the reactor core. Analytical and numerical models were developed and validated by experiments to predict the transmission efficiency of light in the highly polished small-bore light guides. This paper provides details of calculations and experimental validation under laboratory conditions and at a pool-type SLOWPOKE-2 reactor.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Neutron Activation Procedures Used for the Calibration of a Nitrogen-16 Reactor Power Monitor

Personnel from the Pennsylvania State University's Breazeale Nuclear Reactor assisted NRC (Nuclear Research Corporation) personnel in calibrating a new MSL/16N Nitrogen-16 Monitor. This paper describes the neutron flux calibration procedures utilized and the results obtained for the production of a nitrogen-16 source of known activity for a BGO detector calibration.

A Threshold Self-Powered Gamma Detector for Use as a Nuclear Reactor Power Monitor

A study of a threshold self-powered gamma detector for use as a nuclear reactor power monitor was conducted. Measurements were performed to ascertain whether certain detector material arrangements could be used to obtain significant discrimination against low energy gammas. Results indicating agreement between detector response and reactor power output are presented. Evidence of rejection of low energy gammas by the detector is presented. The simplicity of construction and ruggedness of the detector are also discussed.

Gaseous Cerenkov Detector Reactor Power Monitor

A small diameter, gaseous Cerenkov detector has been developed for use as a reactor power monitor. Measurements have confirmed its ability to detect power-related high energy prompt gammas while discriminating against a significant flux of low energy fission product decay gammas. Other advantages of the detector are also mentioned.

Reactor Power Monitor Based on Cherenkov Radiation Detection

At the temperature and radiation levels expected in the core region of some projected reactor plants, conventional radiation sensors, such as fission chambers, suffer severe radiation damage. On the other hand, remote location of sensors introduces geometry effects. These circumstances emphasize the need for a power level sensor which is highly resistant to the hostile environment and still responds directly to power level changes. Such a detection channel has been proposed. It is based on sensing the gamma flux above some energy threshold, thus discriminating against gamma radiation associated with decay. In addition, the proposed system dispenses with cables and voltages applied across damageable insulators, through sensing the light transmitted along a cylindrical duct. The detecting medium is the gas with which the duct is filled; the light signal is the Cherenkov radiation generated by relativistic Compton recoil electrons. Cherenkov radiation intensity increases sharply from an electron velocity threshold, which depends on the index of refraction of the medium. For gases, this threshold can be continuously varied by adjusting the gas pressure. On the other hand, electrons with subthreshold velocities can generate (background) light through scintillation processes. In order to assay the overall discrimination available from the choice of gas, pressure, and geometry, a series of tests has been under way with 3-MeV electron pulses from a Van de Graaff accelerator.

Microwave diagnosis of a plasma generated in a reactor and its use as a reactor power monitor

The plasma produced in a gas filled microwave cavity has been studied by subjecting the cavity to the intense gamma fields present during pulsing operation of a TRIGA nuclear reactor. The cavity studied was designed to serve as an in-core power monitor for extreme power transients and to fulfill the need for a detector of essentially unlimited life. The influence of varying gas pressure and impurity concentration and the effect on sensitivity has been evaluated. High impurity concentrations have been found to greatly influence the cavity performance, reducing free electron densities but improving the response time of the device as a reactor power transient monitor. A prototype device measured gamma ray dose rates from 2 × 10 8 to 2 × 10 10 R/h using natural helium contained in an aluminum cavity. The results are within reasonable agreement with theoretical predictions considering the necessary extrapolation of plasma data.