Reactor Critical Facility Research Articles

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

Inversion Point of the Isothermal Reactivity Coefficient with Increasing System Temperature in a Heterogeneous Core Lattice of a Light Water Reactor

As part of efforts to develop coupled multiphysics experiments for the benchmark of modern multiphysics reactor simulators, a low-power and open-pool type of light water reactor at the Walthousen Reactor Critical Facility (RCF) was reconfigured with additional equipment, and its neutronic characteristics were fully surveyed. A water loop system was designed and installed to pass through the central region of the reactor core, making the central region overmoderated. The overmoderation would lead to a positive temperature reactivity feedback in the modified reactor configuration. This phenomenon is observed when the system temperature is between 10.69°C and 28.70°C. The inversion point of the isothermal reactivity coefficient is at 28.70°C ± 1.07°C. At this temperature, competition between the negative and positive thermal effects on reactivity compensate each other, and the isothermal reactivity coefficient becomes negative at temperatures higher than the inversion point. This paper presents the experimental determination of the isothermal reactivity and reactivity coefficient at different temperatures as well as the inversion point in the modified RCF reactor configuration. To obtain the best-quality results possible, special attention is given to the choice and adaptation of all the available methods for data postprocessing of experiment measurements. Neutron flux denoising is performed with multivariate wavelet transforms and principal component analysis. The Inverse Kinetics Method is applied to derive reactivity from the neutron flux measurements. To provide accurate and high-fidelity experiment benchmark data for modern code validation, in-depth experimental uncertainty quantification is developed. The results of the experiments show the mixed effects of system temperature on reactor reactivity due to the combined effects of Doppler broadening in the fuel, S(α,β) thermal scattering physics, and change in water density and can be used to validate previously developed cross-section interpolation models in the low-temperature range and positive isothermal reactivity coefficient conditions.

Design and fabrication of an axial neutron flux profile measurement assembly for the Advanced Test Reactor Critical Facility

Design and fabrication of an axial neutron flux profile measurement assembly for the Advanced Test Reactor Critical Facility

Evaluation of temperature-dependent critical experiments at the Walthousen reactor critical facility for benchmark development

Development of benchmark libraries based on accurate experimental measurements is crucial for validation of computational toolkits in nuclear reactor design and analysis. A benchmark library is developed based on the low-power isothermal critical experiments conducted at the Walthousen Reactor Critical Facility ranging in temperatures between 16 and 46 Celsius. Sensitivity analysis was performed to evaluate uncertainties in reactivity and its temperature coefficient due to core geometry, material composition, and temperature based on the temperature-dependent experimental data. A comprehensive sensitivity study of the effect of nuclear data on criticality was also performed. The total combined uncertainty in criticality of the benchmark model due to geometry and material parameters is determined to be 144 pcm. Prediction of criticality from sample calculations lies within the benchmark model combined uncertainty from measurements, demonstrating the acceptance of these datasets for use as a benchmark for validation of simulation toolkits and evaluation of nuclear data libraries.

A methodology for uncertainty quantification and sensitivity analysis for responses subject to Monte Carlo uncertainty with application to fuel plate characteristics in the ATRC

Large-scale reactor simulation often requires the use of Monte Carlo calculation techniques to estimate important reactor parameters. One drawback of these Monte Carlo calculation techniques is they inevitably result in some uncertainty in calculated quantities. The present study includes parametric uncertainty quantification (UQ) and sensitivity analysis (SA) on the Advanced Test Reactor Critical (ATRC) facility housed at Idaho National Laboratory (INL) and addresses some complications due to Monte Carlo uncertainty when performing these analyses. This approach for UQ/SA includes consideration of Monte Carlo code uncertainty in computed sensitivities, consideration of uncertainty from directly measured parameters and a comparison of results obtained from brute-force Monte Carlo UQ versus UQ obtained from a surrogate model. These methodologies are applied to the uncertainty and sensitivity of keff for two sets of uncertain parameters involving fuel plate geometry and fuel plate composition.Results indicate that the less computationally-expensive method for uncertainty quantification involving a linear surrogate model provides accurate estimations for keff uncertainty and the Monte Carlo uncertainty in calculated keff values can have a large effect on computed linear model parameters for parameters with low influence on keff.

Patchy nuclear chain reactions

Stochastic fluctuations of the neutron population within a nuclear reactor are typically prevented by operating the core at a sufficient power, since a deterministic (i.e., exactly predictable) behavior of the neutron population is required by automatic safety systems to detect unwanted power excursions. In order to characterize the reactor operating conditions at which the fluctuations vanish, an experiment was designed and took place in 2017 at the Rensselaer Polytechnic Institute Reactor Critical Facility. This experiment however revealed persisting fluctuations and striking patchy spatial patterns in neutron spatial distributions. Here we report these experimental findings, interpret them by a stochastic modeling based on branching random walks, and extend them using a “numerical twin” of the reactor core. Consequences on nuclear safety will be discussed.

Simulated Performance of the Micro-Pocket Fission Detector in the Advanced Test Reactor Critical Facility

The Micro-Pocket Fission Detector (MPFD) is a small-form-factor real-time fission chamber. MPFD performance has been simulated in the Advanced Test Reactor Critical Facility located at Idaho National Laboratory. The neutron and gamma-ray flux profiles and magnitudes were simulated using MCNP in the near-core B-8 irradiation position. These simulations were performed at 69 discrete axial locations inside the B-8 position 55 for three separate orientations of the nearby hafnium outer shim control cylinders and at a power level of 700 W(thermal). The resulting neutron and gamma-ray flux values were used to determine the MPFD response for various fissile masses and detector gas pressures. The optimal gas-operating pressure was determined to be between 30 and 60 psig. The required fissile layer mass was determined to be between 0.5 to 1.0 µg of 235U. Additionally, the gamma ray to fission fragment interaction rate was determined to be 1.43 × 103 with average energy deposition for gamma rays and fission fragments in 30 psig argon gas to be 1 keV and 3.5 MeV, respectively.

Experimental measurements of isothermal reactivity coefficient and temperature-dependent reactivity changes with associated uncertainty evaluations

The low power Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute has recently been utilized to design and perform low-power temperature dependent experiments. The experiments include determination of isothermal excess reactivity and reactivity coefficient at different temperatures and measurements of time-dependent reactivity change with system temperature change. The above-mentioned experiments have produced significant quantities of data, and numerous post-processing steps are needed in order to make the data useful for computational model validation. One important step toward providing the data as an accurate and high-fidelity experiment benchmark for modern code validation is to develop in-depth experimental uncertainty quantification. In this paper, we analyze in particular two types of reactor physics experiments driven by the change in system temperature within a small temperature range. Uncertainty quantification includes separately derived experimental errors for both independent (temperature) and dependent (reactivity) variables, and an iterative weighted least squares approach is used to combine them into the equivalent uncertainty in reactivity. The results of the experiments show the negative effects of the system temperature on reactor reactivity due to combined effects of the Doppler broadening in the fuel, the S(α,β) thermal scattering physics and the decrease in water density. The low uncertainty values obtained for the reactor isothermal excess reactivity (<0.10 cents), the isothermal reactivity coefficient (<0.01 cents/°C), and the time-dependent reactivity change with system temperature increase (<0.16 cents) show that these experiments can be reliably used as benchmarks for code validation. The iterative method adapted in this work for combining temperature and reactivity uncertainties can be used to analyze similar temperature-dependent reactivity benchmark experiments.

Development of critical experiments to provide validation data for multiphysics coupling codes

Modern reactor simulation tools provide advanced prediction capabilities by coupling multiphysics models to simulate reactor behaviors, involving thermal, neutronic, and mechanical interactions. To assure high-fidelity predictions by these tools, experimental data are needed to validate coupled-physics models deployed by these tools. In order to provide data to benchmark the feedback between thermal-hydraulic and neutronic simulations, coupled-physics critical experiments are designed and performed at a Reactor Critical Facility (RCF). The facility houses a low power and open-pool type light water reactor operated at the atmospheric pressure. The reactor allows flexible reconfigurations for many unique critical experiments. Recently, a water loop system has been designed and installed in the facility with the heated water circulating through the center of the reactor core, which broadens the range of validation experiments available for neutronics/thermal-hydraulics couplings. Direct effects of the loop water thermal dynamic change on reactor power and derived reactivity are demonstrated through a series of different experiments, including reactivity change over different loop water temperatures and reactor power evolution under influence of flow transient conditions in the water loop. Changes as low as 1% in the reactor power/neutron flux caused by small water temperature perturbations are observable experimentally.

Development of a research reactor protocol for neutron multiplication measurements

A new series of subcritical measurements has been conducted at the zero-power Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI) using a 3He neutron multiplicity detector. The Critical and Subcritical 0-Power Experiment at Rensselaer (CaSPER) campaign establishes a protocol for advanced subcritical neutron multiplication measurements involving research reactors for validation of neutron multiplication inference techniques, Monte Carlo codes, and associated nuclear data. There has been increased attention and expanded efforts related to subcritical measurements and analyses, and this work provides yet another data set at known reactivity states that can be used in the validation of state-of-the-art Monte Carlo computer simulation tools. The diverse (mass, spatial, spectral) subcritical measurement configurations have been analyzed to produce parameters of interest such as singles rates, doubles rates, and leakage multiplication. MCNP®6.2 was used to simulate the experiment and the resulting simulated data has been compared to the measured results. Comparison of the simulated and measured observables (singles rates, doubles rates, and leakage multiplication) show good agreement. This work builds upon the previous years of collaborative subcritical experiments and outlines a protocol for future subcritical neutron multiplication inference and subcriticality monitoring measurements on pool-type reactor systems.

Upgrading Limiting Peak-Power Analysis Techniques with Modern Validation and Uncertainty Quantification for the Advanced Test Reactor

This work supports the acceptability of the two-dimensional deterministic transport code HELIOS to replace the legacy diffusion code PDQ for computing the peak-power performance parameters of the Advanced Test Reactor (ATR). The 95% Confidence Rule, commonly used in the commercial reactor sector, is explored to develop the so-called reliability factors that provide statistical confidence that the peak-power limits within the hottest location along a fuel plate, referred to as the hot stripe, will not be exceeded. Additionally, an alternative “legacy” methodology was explored that attempts to mimic the exact PDQ analysis process used for defining the peak-power limits. The legacy methodology involves interpolating power between regions at azimuthal boundaries subtending the regions of interest.In order to apply the 95% Confidence Rule, a statistically significant calculation-to-measurement bias must first be established. Unlike the commercial world, where thousands of power observations can be collected every cycle using online flux-mapping instrumentation, the ATR power distribution must be measured during “depressurized” zero-power measurements using fission wires in polyethylene wands. In 2012, fission wire activation data were collected during a flux run in the Advanced Test Reactor Critical Facility. Also to improve statistical validity, archival data from ATR zero-power flux runs from 1977, 1986, and 1994 were digitized from scanned reports and used to create new benchmark models. Borrowing from least-squares adjustment methods commonly used for neutron activation spectroscopy, adjusted fission wire powers were calculated for all four data sets. The mean and standard deviation of the bias between a priori calculated and adjusted wire powers were then taken as the bias and uncertainty used in the 95% Confidence Rule.

Measurement of formation cross-section of 99Mo from the 98Mo(n,γ) and 100Mo(n,2n) reactions

The formation cross-section of medical isotope 99Mo from the 98Mo(n,γ) reaction at the neutron energy of 0.025eV and from the 100Mo(n,2n) reaction at the neutron energies of 11.9 and 15.75MeV have been determined by using activation and off-line γ-ray spectrometric technique. The thermal neutron energy of 0.025eV was used from the reactor critical facility at BARC, Mumbai, whereas the average neutron energies of 11.9 and 15.75MeV were generated using 7Li(p,n) reaction in the Pelletron facility at TIFR, Mumbai. The experimentally determined cross-sections were compared with the evaluated nuclear data libraries of ENDF/B-VII.1, CENDL-3.1, JENDL-4.0 and JEFF-3.2 and are found to be in close agreement. The 100Mo(n,2n)99Mo reaction cross-sections were also calculated theoretically by using TALYS-1.8 and EMPIRE-3.2 computer codes and compared with the experimental data.

In-Core Flux Sensor Evaluations at the ATR Critical Facility

As part of an Idaho State University (ISU)–led Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) collaborative project that includes Idaho National Laboratory (INL) and the French Alternative Energies and Atomic Energy Commission (CEA), flux detector evaluations were completed to compare their accuracy, response time, and long-duration performance. Special fixturing, developed by INL, allows real-time flux detectors to be inserted into various Advanced Test Reactor Critical Facility (ATRC) core positions to perform lobe power measurements, axial flux profile measurements, and detector cross-calibrations. Detectors initially evaluated in this program included miniature fission chambers, specialized self-powered neutron detectors (SPNDs), and specially developed commercial SPNDs. Results from this program provide important insights related to flux detector accuracy and resolution for subsequent ATR and CEA experiments and yield new flux data required for benchmarking models in the ATR Life Extension Program (LEP) Modeling Update Project.

Nondestructive Assay Measurements Using the RPI Lead Slowing-Down Spectrometer

The use of a lead slowing-down spectrometer (LSDS) is considered as a possible option for nondestructive assay of fissile material of used nuclear fuel. The primary objective is to quantify fissile isotopes, particularly 239Pu and 235U, via a direct measurement distinguishing them through their characteristic fission spectra in the LSDS. In this paper, we present several assay measurements performed at the Rensselaer Polytechnic Institute (RPI) to support ongoing feasibility studies of the method and to provide benchmark experiments for Monte Carlo calculations of the assay system. A fresh uranium oxide fuel rod from the RPI Walthousen Reactor Critical Facility, a 239Pu-Be source, and several highly enriched 235U disks were assayed in the LSDS. The characteristic fission spectra were measured with 238U and 232Th threshold fission chambers, which are primarily sensitive to fission neutrons with energies above the threshold. Despite the constant neutron and gamma background from the Pu-Be source and the intense interrogation neutron flux, the LSDS system was able to measure the characteristic 235U and 239Pu responses. All measurements were compared to Monte Carlo simulations complementing previous modeling-based studies. It is shown that the available simulation tools and models are well suited to simulate the assay. An absolute calibration technique of the LSDS, which is required to perform quantitative measurements of the assayed fissile materials, is presented.

Large sample neutron activation analysis of dross for gold and silver

Large sample neutron activation analysis of dross from India Government Mint, Mumbai was carried out for quantification of gold (Au) and silver (Ag) using graphite reflector position of Advanced Heavy Water Reactor critical facility at Bhabha Atomic Research Centre, Mumbai. The k0-based internal monostandard NAA was used to calculate concentration ratios of Au and Ag with respect to sodium (Na), which was used as an internal monostandard. The concentration ratio values of Au to Na of varying mass of dross showed that mass ≥2 g was the representative sample size for analysis. Concentrations of gold and silver were found to be in the range of 200–400 and 1200–1700 mg kg−1, respectively in three different samples.

Measurements of Delayed Fission Product Gamma-Ray Transmission Through Low-Enriched UO2 Fuel Pin Lattices in Air

Measurements of delayed fission product gamma-ray transmission through low-enriched UO2 fuel pin lattices in an air medium were conducted at the Rensselaer Polytechnic Institute Reactor Critical Facility (RCF). The RCF core consists of excess Special Power Excursion Reactor Test fuel pins enriched to 4.81 wt% 235U and clad in stainless steel. An experimental apparatus was constructed to hold various arrangements of fuel pin lattices. The arrangements consisted of a single activated source pin taken from the reactor core surrounded by inactive fuel pins in an air medium. A sodium-iodide detector and gamma-ray spectroscopy system was used to generate a pulse-height spectrum of the gamma-ray radiation for detector positions outside the lattice. The change in radiation intensity as the detector is rotated about the vertical axis of the lattice, the “channeling effect,” was measured. Measurements of the channeling effect were performed for six experimental arrangements: 3 × 3, 5 × 5, and 7 × 7 lattices, with both the corner and the center positions containing the activated pin. The results of the measurements demonstrate that the gamma-ray radiation intensity can vary widely as a function of angle relative to the angle of rotation about the vertical axis of the lattice.

Funding and Operating University Reactors

On page 23 of its April 2002 issue, Physics Today reviewed the perilous state of university reactors. That article contained several oversights that may be of interest to the physics community.First, most university reactors can be characterized as small “neutron factories” that produce neutrons for neutron scattering and activation measurements. Reactor use in those applications depends largely on faculty interest and research funding, which in some cases have been marginal. 1 1. M. Corradini, M. Adams, D. Dei, T. Isaacs, G. Knoll, Warren Miller, K. Rogers, The Future of University Nuclear Engineering Programs and University Research & Training Reactors (rep. prepared for the US Department of Energy Office of Nuclear Energy), DOE, Washington, DC (May 2000). Final draft available online at http://www.nuclear.gov/nerac/finalblue.pdf. Moreover, neutron scattering and activation experiments may not be closely related to the nation’s nuclear power needs.One university reactor, however—operated by Rensselaer Polytechnic Institute in Troy, New York—is directed at reactor physics measurements and at training that is directly pertinent to the nation’s pressurized water reactors (PWRs) and boiling water reactors (BWRs). The RPI Reactor Critical Facility is alone among university reactors in that it uses fuel that is very much like commercial fuel—that is, high-density uranium oxide pellets enriched to 4.81 weight percent uranium-235 and stacked in metal cladding. The RCF measurements are similar to the “startup measurements” that every power reactor is required to make each year or two. The measurements include (1) the effects of changes in state variables such as temperature and voids on multiplication factor and (2) the effects of core structure and composition on power shape and multiplication factor.Second, most RCF measurements are made at low power, close to 1 W. In fact, to permit access to the core for experiments, one does not want to activate the fuel by going to higher power. The annual operating energy production is usually less than .05 kWh/year, with the result that radiation doses are negligible and the facility is not appreciably activated. Another consequence is that the operating budget has been less than $40 000 per year, most of which is paid by tuition and grants. That figure is much less than the costs of $100 000 and more that may be “shouldered” by the universities.Third, some university reactors are characterized and valued as training reactors in that they permit “real” reactor experience that contributes to learning a “safety culture.” That opinion is not universally held. Reactor simulators provide a very realistic training experience for power reactor operation. Nuclear power plants almost exclusively use simulators for training purposes, whereas they formerly used training reactors (including that at RPI) as well.REFERENCESection:ChooseTop of pageREFERENCE <<1. M. Corradini, M. Adams, D. Dei, T. Isaacs, G. Knoll, Warren Miller, K. Rogers, The Future of University Nuclear Engineering Programs and University Research & Training Reactors (rep. prepared for the US Department of Energy Office of Nuclear Energy), DOE, Washington, DC (May 2000). Final draft available online at http://www.nuclear.gov/nerac/finalblue.pdf. Google Scholar© 2003 American Institute of Physics.

Experimental Studies of Radially Heterogeneous Liquid-Metal Fast Breeder Reactor Critical Assemblies at the Zero-Power Plutonium Reactor

Experimental programs to investigate the physics characteristics of heterogeneous liquid metal fast breeder reactor cores have been conducted in the zero-power plutonium reactor critical facility over a period of about5 yr. Previous experiments on conventional homogeneous cores provided appropriate benchmark data against which to judge the heterogeneous core results. For a heterogeneous reactor of the Clinch River Breeder Reactor size, both the physics parameters and the ability to predict them by common design methods differ substantially from an equivalent conventional design. Data errors and methods approximations have a greater effect in the analysis of heterogeneous cores, particularly with respect to such spatially varying parameters as power distributions and control rod worths. Preliminary results from recent experiments on a 700-MW (electric)-sized heterogeneous assembly are presented. As expected, predictions of physics parameters in general are worse than for conventional cores. Eigenvalue spectra and cross-section sensitivity have been used to characterize the spatial sensitivity of the cores.