CROCUS Reactor Research Articles

High-Resolution Modeling Without Computation Slowdown for PETALE in CROCUS

In a collaboration between Ecole Polytechnique Fédérale de Lausanne (EPFL) and CEA, in the fall of 2020, the experimental Programme d’Étude en Transmission de l’Acier Lourd et ses Eléments (PETALE) was successfully carried out in the CROCUS reactor of EPFL. This article presents and compares the methods tested in the modeling of the experiments, specifically focusing on the metal reflectors installed at the periphery of CROCUS. A basic design model consisting of a few cuboids was refined to a fully detailed version, without impacting the run time of simulations. Notably, each reflector sheet of PETALE was segmented into 121 voxels based on topological measurements. This detailed voxelization did not affect calculation times, thanks to the use of three-dimensional lattices as available in Serpent 2. Profiling of the simulations revealed the high computational surface transformations associated with Serpent 2 and highlighted the efficiency benefits of factorizing these into universe transformations. As the CROCUS simulations were carried out using a modified build of Serpent 2, additional simulations were also performed using a standard version of Serpent 2 with a GODIVA model as a neutron source to ensure that the findings are generalizable. These additional tests confirmed the initial results, with significant performance variations observed between the models, particularly larger in surface-tracking mode than in delta-tracking mode. Consequently, the modeling method may therefore be applied to future high-fidelity modeling of neutron transmission and shielding experiments.

Fast simulation of neutron noise using the Transient Fission Matrix approach and validation on the CROCUS reactor

This article presents the development of a neutron noise calculation technique with a reduced calculation time based on the TFM (Transient Fission Matrix) approach. The latter contains different information such as the system transfer function (fission and absorption probability density functions and corresponding neutron multiplicity) of prompt and delayed neutrons, and the corresponding propagation time. This information is used to reconstruct neutron showers based on precursor decays up to the shower disappearance by absorption, leakage or delayed neutron precursor creation. Then pseudo-detectors such as fission chamber count rate histograms are reconstructed from these neutron showers and the αprompt is calculated from the cross-correlation function. A very good agreement has been obtained on the uranium-fueled light water reactor CROCUS between the model developed and the reference value.

Perovskite-based detector for reactor dosimetry monitoring

Hybrid halide perovskites have demonstrated significant efficiency in detecting a broad spectrum of high-energy radiation, including X-rays, gamma rays (γ-rays), and neutrons. Given the common occurrence of mixed radiation fields, we investigated the performance of a perovskite-based detector in a neutron-gamma mixed field. A large methylammonium lead tribromide (MAPbBr3) single crystal (SC) was synthesized via the oriented crystal-crystal intergrowth method. This SC was used to fabricate a gamma detector with carbon electrodes, which was tested in the CROCUS zero-power reactor cavity. The detector’s photocurrent response exhibited a strong correlation with known gamma dose rates, as measured by an ambient Berthold LB 112 gamma probe, facilitating the accurate conversion of photocurrent to dose rate. Notably, the device did not exhibit degradation under neutron radiation exposure. To further assess the impact of neutrons, X-ray diffraction and electron paramagnetic resonance analyses were performed on small MAPbBr3 SCs grown by inverse temperature crystallization. These SCs were irradiated within the CROCUS reactor core and by a Pu-Be neutron source at liquid nitrogen temperature. Our findings indicate that the perovskite material can withstand the nominal in-core operation conditions of the CROCUS reactor. Additionally, it endures irradiation at liquid nitrogen temperature, corresponding to a fast neutron fluence of approximately 1010 cm-2 and a gamma radiation dose of about 50 Gy, confirming only the temporary creation of defects. No signs of long-term deterioration were observed, suggesting a potential self-healing mechanism. This resilience positions perovskite SCs as viable candidates for in-core radiation detection, supporting the further development of miniaturized MAPbBr3 SC devices for such applications.

POLLEN: A Pile-Oscillator for the BLOOM Experimental Program

In the scope of its project on the assimilation and reproduction of experiments for the evaluation of stainless-steel nuclear data HARVEST-X, the LRS launches a pile-oscillation experimental program in the CROCUS reactor: BLOOM. For this program, an oscillator called POLLEN originally developed to be used as a vibrating absorber to compensate the fuel rods oscillator COLIBRI was reworked as a standalone pile-oscillator called. The oscillator operates with an arbitrary periodical shape, an amplitude of 1 m and a frequency ranging from the mHz to the Hz. An emphasis was put in the development of an interface for pseudo-square oscillations for BLOOM. This interface also allows independent adjustment of the ramping time and dwell time of the pseudo-square. The qualification of POLLEN was performed by video analysis of sinusoidal oscillations, using a 4K 30fps camera. With the current system, a precision of 0.2 mm was achieved during slow sinusoidal oscillation and a precision of 0.17 mm was obtained for pseudo square oscillations with a 500 g load. Whereas the results are satisfactory with respect to the current system and fulfill the requirement of the BLOOM program, it is planned to upgrade the system with the addition of a mechanical reference in the system. It is also in consideration to upgrade the acquisition card and controller to 16 bits systems to allow the use of POLLEN in setups where larger displacements are needed. The first in-core oscillation experiments are planned for early 2024.

Analysis of the preliminary campaign for the PETALE program

The PETALE program aims to provide new experimental data to constrain the stainless steel nuclear data. In this frame, a preliminary measurement campaign has been performed to characterize the neutron flux in key positions of the CROCUS reactor and to develop analysis tools. During this preliminary campaign detailed in the present paper, an efficiency ratio technique has been developed and tested to speed up HPGe measurements by a factor of 30. A second objective of the campaign concerns the propagation of nuclear data uncertainty from the core neutron cross-sections to the reaction rates in the dosimeters. Uncertainties in the core cross sections, such as the uranium cross section, are nuisance parameters that add uncertainty to the dosimeter reaction rate calculation. This component must be fully characterized with covariances to constrain the metal reflector component for Bayesian assimilation. The experimental results are compared to the calculations with different nuclear databases for the nuclear data uncertainty propagation. A good agreement is obtained with the ENDF/B-VII.1 database and a systematic underestimation of around 5–10% in the fast range is observed with the ENDF/B-VIII.0 and JEFF-3.3 databases.

Characterization of High Harmonic Frequencies in Reactor Noise Experiments Within the CORTEX Project

We present a novel technique of neutron noise detection and experimental data interpretation developed during the EU H2020 project CORTEX aiming to improve the capabilities for identification and localization of neutron noise sources. The experimental data analysis is performed in the frequency domain by extracting the spectral power density and the phase angle using a novel spectral variance reduction technique based on per cycle-based bootstrapping with replacement. This technique allows for variance reduction of measured spectral power and phase angle not only at base frequency, but at higher harmonic frequency contributions as well. This allows for a more representative treatment of experimental data and validation of codes for neutron noise propagation, some of which have been developed within the project. The detector response is not necessarily linearly dependent on the oscillator movement, and the study of nonlinear terms provides additional information which can improve the accuracy of neutron noise source identification and localization. Moreover, these terms can be used in a Taylor series to identify a more complex dependence. The process is simplified in the sense that these contributions are linearized in the frequency domain as higher harmonic frequency contributions and can be easily identified and extracted due to spectral peaks prominence provided by the bootstrapping with replacement method. Combined with the spectral power and phase, we present a preliminary investigation of usability of higher-order terms for noise source identification and localization. In this article, we outline the CORTEX project, the experiments, and the measurement analysis methodology based on the bootstrapping with replacement along with the initial developments on the study of nonlinear terms in CROCUS reactor, using several different noise source configurations at a set frequency.

Development and validation of a time-dependent deterministic model for neutron noise on CROCUS experimental measurements

In the framework of the European project CORTEX, included in the H2020 program, the experimental campaign carried out with the CROCUS reactor, at the École Polytechnique Fédérale de Lausanne (EPFL) in Switzerland, aims at setting up methodologies and tools for the analysis and the interpretation of neutron noise in view of their industrial applications. An important part of the CORTEX project deals with code development and validation. In this paper, a new methodology for the simulation of neutron noise has been formulated and implemented in the APOLLO3 code. In the first part of this paper, the experimental setup and the methodology for the experimental data analysis developed by CEA are presented. The experimental outcomes are reported in terms of Cross-Power Spectral Densities (CPSD) of detector responses. The results produced with CEA analysis are compared to those obtained by EPFL. The second part of the work presents the noise simulator based on an improved point-kinetics model employed in a fully heterogeneous 2D transport calculation including its theoretical derivation. The computed results, i.e. the temporal signals of the detector responses, are treated to recover the CPSDs to be compared to the experimental ones. In the third part of the work, the paper presents a series of interpretive exercises performed with the IPK noise model aiming at showing its simulation capabilities and at trying to address some of the discrepancies observed in the direct comparison. Beside the promising outcomes of the simulations, some work is still being done from the experimental and computational side for a better comprehension of the problem before the extension of the model to full 3D problems and industrial application.

Local and high distance neutron and gamma measurements of fuel rods oscillation experiments

We report in the present article on the successful observation using noise analysis of the lateral oscillation of one fuel rod by ±2.5 mm around nominal at 0.1 Hz frequency, using an mm3miniature neutron scintillatorat the rod level, and a BGO gamma detectorseven meters away from the reactor core center. The experiment was conducted as part of the COLIBRI program in the CROCUS reactor, which is dedicated to the investigation of reactor noise induced by fuel vibrations. It consists in experiments on rod lateral displacement (static) and oscillation (dynamic) with different rods’ numbers at various relevant amplitudes and frequencies. Its main motivation is the increased amplitudes in the neutron noise distributions recorded in ex- and in-core detectors that have been observed in recent years in Siemens pre-Konvoi type of PWR reactors. The obtained experimental data are used for the purpose of code validation, especially within the framework of the European project CORTEX on reactor noise applications. During the first phase of COLIBRI, the observation of a spatial dependence of the perturbation noise, also called neutron modulation, was demonstrated. In the second phase of COLIBRI starting 2021, it is planned to use a core mapping array of neutron detectors to record its propagation. It consists in about 150 miniature scintillators coupled to optical fibers and SiPM readouts, to be distributed in the reactor core. As a feasibility test, experiments were performed using a miniature scintillator prototype placed on a fuel rod, and oscillating the instrumented rod or the one directly adjacent to the detector. In addition, it is theoretically possible to measure branching or perturbation reactor noise using gamma radiation. Following recent developments on gamma measurements in CROCUS, the fuel oscillation was simultaneously recorded with a gamma detection array, LEAF. Its large BGO detectors were used by placing them at the maximum distance to the core, i.e. seven meters away with a clear line of sight using an experimental channel through the shielding of the reactor cavity.

CORE SIM+ SIMULATIONS OF COLIBRI FUEL RODS OSCILLATION EXPERIMENTS AND COMPARISON WITH MEASUREMENTS

At EPFL, the CROCUS reactor has been used to carry out experiments with vibrating fuel rods. The paper presents a first attempt to employ the measured data to validate CORE SIM+, a neutron noise solver developed at Chalmers University of Technology. For this purpose, the original experimental data are processed in order to extract the necessary information. In particular, detector recordings are scrutinized and detrended, and used to estimate CPSDs of detector pairs. These values are then compared with the ones derived from the CORE SIM+ simulations of the experiments. The main trend of the experimental data along with the values for some detectors are successfully reproduced by CORE SIM+. Further work is necessary on both the experimental and computational sides in order to improve the validation process.

PISTIL, a reactivity modulation device to probe the transfer function of the nuclear reactor CROCUS

The present article summarizes the development and testing of a reactivity modulation device developed by the French Atomic and Alternative Energies Commission (CEA). It was installed in the CROCUS reactor of the Swiss Federal Institute of Technology in Lausanne (EPFL). Experimental tests were performed in the framework of a collaboration between CEA and EPFL. The so-called PISTIL device aims at measuring the nuclear reactor transfer function in the frequency range of interest between 1 mHz and 200 Hz, in order to probe the in-core kinetic behavior of prompt and delayed neutrons. The reactivity modulation is obtained from the rotation of cadmium foils. The design of the system was driven with the objective of installing PISTIL at the center of the CROCUS reactor. Neutronic simulations with TRIPOLI-4 Monte Carlo code were performed to select the suitable design parameters and meet the safety requirements of the reactor operation. The total reactivity worth of the device, as estimated by TRIPOLI-4 Monte Carlo calculation, was approximately 0.16 $ and the maximum amplitude of the reactivity modulation was about 0.013 $. In-core reactivity calibration was then performed and were consistent as compared to TRIPOLI-4 estimations.

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 CeBr3scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6±0.6)%.

Design of a 150-miniature detectors 3D core-mapping system for the CROCUS reactor

The present article provides an overview of the design of a three-dimensional (3D) full-core mapping system for the CROCUS reactor, operated at the École polytechnique fédérale de Lausanne (EPFL), Switzerland. The system is composed of 149 miniature neutron detectors distributed within the core double lattice at three main axial levels. The miniature detector technology is based on the optimization of the well-proven coupling of a miniature ZnS:6LiF(Ag) scintillator to a state-of-the-art silicon photomultiplier (SiPM) via jacketed optical fibers. The challenges in the mechanical design, the detector optimization, the core criticality, and the development of the acquisition electronics are strongly interconnected and their combination is addressed in this article. The 3D full-core mapping system is foreseen to be installed in CROCUS in autumn 2021 and it will pave the way for the investigation of 3D dynamic phenomena in nuclear reactor cores.

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.

HIGHLY LOCALIZED AZIMUTHAL MEASUREMENTS IN THE CROCUS REACTOR TOWARDS THE VALIDATION OF HIGH-FIDELITY NEUTRONICS CODES

Highly localized in-core measurements are necessary for the validation of neutron transport calculations with high spatial resolution. In the present work, a miniature neutron detector developed at EPFL in collaboration with PSI was used to carry out a set of thermal neutrons counting measurements in the zero-power CROCUS reactor core within a spatial range in order of mm. The miniature detector, positioned close to the core reflector, shows a gradient of +(4.29 ± 0.10)% in the count rate profile in the radial direction within 1.3 cm, with higher values pointing towards the core reflector because of the higher share of neutrons in the thermal range. On the contrary, in a control rod guide tube the count rate gradient is -(4.37 ± 0.10)% and it is directed towards the core center. The measured values are compared with the azimuthal trend of the normalized 6Li reaction rate calculated with an iterative three-steps method performed with the Monte Carlo code Serpent 2. These measurements proved the feasibility of resolving spatial effects in the mm-range and they represent a basis for further investigating highly spatially-resolved phenomena in the CROCUS core.

Analysis of the first COLIBRI fuel rods oscillation campaign in the CROCUS reactor for the European project CORTEX

Analysis of the first COLIBRI fuel rods oscillation campaign in the CROCUS reactor for the European project CORTEX

A mm3 Fiber-Coupled Scintillator for In-Core Thermal Neutron Detection in CROCUS

This article presents the development, full-characterization, and in-core testing of a miniature neutron detector for the investigation of highly localized in-core neutronics effects in the zero-power reactor CROCUS, operated at the Ecole Polytechnique federale de Lausanne (EPFL), Switzerland. A ZnS: 6LiF screen, mixing inorganic scintillator and neutron converter, with a surface of 1 mm2 and a thickness of 0.2 mm, was coupled with a silicon photomultiplier (SiPM) via a 10-m optical fiber. Analog readout electronics were adopted in this prototype version to process the SiPM output signal. A first campaign was carried out to assess the detection system capabilities in terms of neutron detection and counting in mixed neutron-gamma radiation fields. The miniature detector was thoroughly characterized with a Pu-Be source installed in the CARROUSEL facility and was subsequently tested inside a control rod guide tube of the CROCUS reactor in different reactor conditions, i.e., at shutdown, startup, and critical. The detector shows a linear response with a reactor power increase up to 6.5 W (i.e., around $10 {^{\mathrm{ 8}}}\,\,\text {cm}^{-2}\cdot \text {s}^{-1} $ total neutron flux) and excellent neutron counting capabilities, when compared to other localized detection systems available in CROCUS, such as the miniature fission chambers and single-crystal vapor deposited (sCVD) diamond detectors. In addition, the exposure of the miniature detector to a 60Co source with an activity of ~250 GBq, combined with the computation of the gamma flux in CROCUS with the Serpent 2 Monte Carlo code, confirmed the gamma insensitivity of the system in a mixed neutron-gamma field.

Design and Simulation of Gamma Spectrometry Experiments in the CROCUS Reactor

Gamma rays in nuclear reactors, arising either from fission 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 computed spectra, dose estimates, reactor noise and prompt to delayed gamma ratios. This data then contributes to models, code validation and photo atomic nuclear data evaluation. In order to contribute to aforementioned experimental data, gamma detection capabilities are being added to the CROCUS reactor facility. The CROCUS reactor 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). With a maximum power of 100W, it is a zero power reactor used for teaching and research, most recently for intrinsic and induced neutron noise studies. For future gamma detection applications in the CROCUS reactor, an array of four detectors - two large 5”x10” Bismuth Germanate (BGO) and two smaller Cerium Bromide (CeBr3) scintillators - was acquired. The BGO detectors are to be arbitrarily positioned in the core reflector and out of the vessel for measurements at arbitrary distances. The CeBr3 detectors on the other hand are small enough to be set in the guide tubes of the control rods for in-core measurements. We present a study of the neutron and gamma flux in the core and reflector using the MCNP 6.2 and Serpent 2 Monte Carlo codes for coupled neutron and photon transport criticality calculations. More specifically, we investigate and compare predicted spectra as well as reactivity worth of different envisioned experimental setups. We further predict pulse height spectra as well as doses to the crystals with and without cadmium shielding to estimate allowable reactor powers with respect to detector radiation hardness. The results serve as basis for calibration and aid in the design and regulatory approval of the experiments.

In-core dosimetry for the validation of neutron spectra in the CROCUS reactor

The present article describes the preliminary validation study of simulated in-core and reflector n eutron spectra in preparation of oncoming experimental programs in the zeropower reactor CROCUS at EPFL. For this purpose, a set of activation foils were irradiated at three characteristic positions in the CROCUS reactor, and the subsequent activities were analyzed via γ spectrometry. The experimental setup was then modeled with the Monte Carlo neutron transport code Serpent2 and associated with an analysis tool to include the effect of the reactor power history during experiments. The comparison of calculated and measured reaction rates (C/E) indicates a general consistency (at 2σ) between calculated and measured spectra. However, offsets of C/E values were observed in (n, γ) reactions, up to 18% for 115In and 8% for 63Cu dosimeters. This could be caused by an unexpected isotopic composition, uncertainties in nuclear data, or the spectrometry analysis. In addition, a 100-groups spectrum unfolding was performed using the experimentally determined reaction rates and the Serpent2 spectra as the prior knowledge. The unfolded spectra were mainly adjusted in the thermal and fast ranges, while few modifications w ere m ade i n t he e pithermal r egion d ue t o the low contribution of epithermal neutrons in activation processes. Moreover, within energy groups where the capture reactions show resonant behavior, flux depletion (up to 38% as compared to the prior spectra) is observed due to the absence of self-shielding effect in the unfolding process. For this purpose, an unfolding method based on energy groups weighting is developed and tested.

Power calibration methodology at the CROCUS reactor

In the present article, we detail the method used to experimentally determine the power of the CROCUS zero-power reactor, and to subsequently calibrate its ex-core monitor fission chambers. Knowledge of the reactor power is a mandatory quantity for a safe operation. Furthermore, most experimental research programs rely on absolute fission rates in design and interpretation – for instance, tally normalization of reaction rate studies in dosimetry, or normalization of power spectral density in neutron noise measurements. The minimization of associated uncertainties is only achieved by an accurate power determination method. The main experiment consists in the irradiation, and therefore, the activation of several axially distributed Au-197 foils in the central axis of the core, which activities are measured with a High-Purity Germanium (HPGe) gamma spectrometer. The effective cross sections are determined by MCNP and Serpent Monte Carlo simulations. We quantify the reaction rate of each gold foil, and derive the corresponding fission rate in the reactor. The variance weighted average over the distributed foils then provides a calibration factor for the count rates measured in the fission chambers during the irradiation. We detail the calibration process with minimization of respective uncertainties arising from each sub-step, from power control after reactivity insertion, to the calibration of the HPGe gamma spectrometer. Biases arising from different nuclear data choices are also discussed.