Gamma noise to non-invasively monitor nuclear research reactors

Autonomous nuclear reactor monitoring is a key aspect of the International Atomic Energy Agency's strategy to ensure nonproliferation treaty compliance. From the rise of small modular reactor technology, decentralized nuclear reactor fleets may strain the capacities of such monitoring and requires new approaches. We demonstrate the superior capabilities of a gamma detection system to monitor the criticality of a zero power nuclear reactor from beyond typical vessel boundaries, offering a powerful alternative to neutron-based systems by providing direct information on fission chain propagation. Using the case example of the research reactor CROCUS, we demonstrate how two bismuth germanate scintillators placed outside the reactor vessel can precisely observe reactor criticality using so called noise methods and provide core status information in seconds. Our results indicate a wide range of applications due to the newly gained geometric flexibility that could find use in fields beyond nuclear safety.

First in-core gamma spectroscopy experiments in a zero power reactor

Gamma rays in nuclear reactors, arising either from nuclear reactions or decay processes, significantly contribute to the heating and dose of the reactor components. Zero power research reactors offer the possibility to measure gamma rays in a purely neutronic environment, allowing for validation experiments of dose estimates, computed spectra, and prompt to delayed gamma ratios. The resulting data can contribute to models, code validation and photo atomic/nuclear data evaluation. To date, most experiments have relied on flux measurements using TLDs, ionization chambers, or spectrometers set in low flux areas. The CROCUS reactor allows for flexible detector placement in and around the core, and has recently been outfitted with gamma detection capabilities to fulfill the need for in-core gamma spectroscopy, as opposed to flux. In this paper we report on the experiments and accompanying simulations of gamma spectrum measurements inside a zero power reactor core, CROCUS. It is a two-zone, uranium-fueled light water moderated facility operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology Lausanne (EPFL). Herein we also introduce, in detail, the new LEAF system: A Large Energy-resolving detection Array for Fission gammas. It consists of an array of four detectors – two large ø 127 254 mm Bismuth Germanate (BGO) and two smaller ø 12 50 mm Cerium Bromide (CeBr3) scintillators. We describe the calibration and characterization of LEAF followed by first in-core measurements of gamma ray spectra in a zero power reactor at different sub-critical and critical states, and different locations. The spectra are then compared to code results, namely MCNP6.2 pulse height tallies. We were able to distinguish prompt processes and delayed peaks from decay databases. We present thus experimental data from hitherto inaccessible core regions. We provide the data as validation means for codes that attempt to model these processes for energies up to 10 MeV. We finally draw conclusions and discuss the future uses of LEAF. The results indicate the possibility of isotope tracking and burn-up validation.

Delayed gamma fraction determination in the zero power reactor CROCUS

Gamma rays are an inextricable part of a nuclear reactor’s radiation field, and as such require characterization for dose rate estimations required for the radiation protection of personnel, material choices, and the design of nuclear facilities. Most commonplace radiation transport codes used for shielding calculations only included the prompt neutron induced component of the emitted gamma rays. The relative amount of gamma rays that are emitted from delayed processes – the delayed gamma fraction – amount to a significant contribution, e.g. in a typical zero power reactor at steady state is estimated to be roughly a third. Accurate predictions of gamma fields thus require an estimation of the delayed content in order to meaningfully contribute. As a consequence, recent code developments also include delayed gamma sources and require validation data. The CROCUS zero power research reactor at EPFL is part of the NEA IRPhE and has therefore been characterized for benchmark quality experiments. In order to provide the means for delayed gamma validation, a dedicated experimental campaign was conducted in the CROCUS reactor using its newly developed gamma detection capabilities based on scintillators. In this paper we present the experimental determination of the delayed gamma fraction in CROCUS using in-core neutron and gamma detectors in a benchmark reactor configuration. A consistent and flexibly applicable methodology on how to estimate the delayed gamma fraction in zero power reactors has hitherto not existed – we herein present a general experimental setup and analysis technique that can be applied to other facilities. We found that the build-up time of relevant short lived delayed gamma emitters is likely attributed to the activation of the aluminium cladding of the fuel. Using a CeBr3 scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6±0.6)%.

DOSIMETRY MODELING AND EXPERIMENTAL VALIDATION FOR THE PETALE PROGRAM IN THE CROCUS REACTOR

The PETALE experimental program in the CROCUS reactor intends to provide integral measurements on reactivity worth and dosimetry measurement to constrain nuclear data relative to stainless steel heavy reflectors. The experimental setup consists in eight successive plates of pure iron, pure nickel, pure chromium, or nuclear-grade stainless steel set at the close periphery of the core. The plates are interleaved with up to nine dosimeters that consist of thin activation foils with different possible materials to be sensitive to different ranges of the neutron spectrum. A precise measurement with a good estimation of the uncertainties and correlations is required, especially when comparing reaction rates, e.g. transmission measurement and/or spectral indices.

The present work focuses on the validation of the dosimetry technics developed in preparation of this experimental program. Different aspects are discussed: monitors, efficiency calibration, self-absorption correction, self-shielding and nuclear data uncertainties. The different sources of uncertainties for the experiment-calculation comparisons are characterized, taking into account all the correlation between the different dosimeters. These correlations are a mandatory element for the aimed Bayesian assimilation in order to avoir overfitting when considering dosimeter providing a similar information.