Experimental measurements of isothermal reactivity coefficient and temperature-dependent reactivity changes with associated uncertainty evaluations
Not provided.
Nuclear Science & Technology↗
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Validation of Jezebel Reactivity Coefficients and Sensitivity Analysis
Nuclear data validation is often performed today using criticality measurements. The gold standard for criticality measurements is the International Criticality Safety Benchmark Experiment Project (ICSBEP). The validation specifically focuses on the effective multiplication factor ($k_{eff}$). $K_{eff}$ is a relatively easy parameter to infer and has reduced uncertainty due to being at or above critical. However, while $k_{eff}$ is the most documented parameter and its uncertainties and sensitivities have been evaluated in great detail, it cannot be used as a standalone metric to determine inaccuracies in nuclear data (e.g., cross section data, PFNS, nu), which is based on theory, physics, and differential measurements. The Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) project aims to identify compensating errors in specific isotope nuclear data by optimally designing experiments that are, or are not sensitive to a suite of measurement parameters beyond $k_{eff}$. By identifying parameters that are sensitive to each other, oppositely sensitive, or have substantial magnitude differences in sensitivity, experiments can be designed to constrain questionable nuclear data. One sensitivity that is of particular interest to this project includes the sensitivity of reactivity coefficients. Reactivity coefficients compare reactivity, which is related to $k_{eff}$ at two different states therefore being sensitive to small changes in the system. The most common type of reactivity coefficient measurements is comparison to void for a small sample within the assembly. It is key that the sample sizes are small enough to not affect the flux of the full system. Reactivity coefficients were evaluated for many early experiments to better understand transport corrected cross sections. In fact, ICSBEP includes reactivity coefficient results as “Supplemental Measurements” in appendices for a handful of older benchmarks. One of those benchmarks is Jezebel, the bare Pu critical assembly. This paper compares new simulations of reactivity coefficients for Jezebel, and explores the sensitivity of reactivity coefficients to small changes in nuclear data.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Reactivity Coefficient Calculation for a Low-Power LEU Research Reactor
Core design studies on a 20-MW thermal power low enriched uranium (LEU) fueled research reactor are underway at the NIST Center for Neutron Research (NCNR) with the primary purpose of providing quality neutron beams for scientific research. The 'tank-in-pool' design pattern is adopted for the new reactor: A compact core composed of two horizontally split halves are surrounded by a cylindrical D{sub 2}O reflector tank, and the tank is itself surrounded by a large H{sub 2}O pool, which functions as thermal and biological shielding. The split core consists of 18 MTR-type fuel elements which are evenly distributed into two regions. The LEU fuel used in the study is the U{sub 3}Si{sub 2}-Al dispersion fuel with U-235 enrichment of 19.75%, the only NRC certified LEU fuel in the U.S. The core is cooled and moderated by H{sub 2}O, and separated from D{sub 2}O by zircaloy core boxes. Hafnium control blades are utilized as both criticality and safety control elements for the reactor. They are controlled by a mechanical driver located at the bottom of the core but with the fully withdrawn positions at the top of the core. In a standard reactor calculation, it is necessary to evaluate the reactivity coefficient with respect to some physical parameters such as fuel or moderator temperature. For nearly all reactor designs, negative reactivity coefficients are required for safety. As a light-water cooled thermal reactor, the new reactor is designed to work in an under-moderated mode with the most achievable reactivity. The core is required to have negative reactivity coefficients under any normal or abnormal circumstances such as over power transient, loss of coolant or reflector, etc. Therefore the reactivity coefficient addresses reactor safety concerns as well as provides valuable quantitative parameters for reactor safety analyses. In this summary, three typical reactivity coefficients for the low-power LEU core are calculated using the Monte-Carlo code MCNP-6: the moderator temperature coefficient (MTC), the fuel temperature coefficient (FTC), and the void coefficient (VC). A direct perturbation methodology is applied to calculate all these coefficients. The perturbed case is produced by manually perturbing a single physical property with respect to the reference case (the one with normal operational conditions), as the multiplication factor (i.e. k{sub eff}) of the perturbed case is calculated by MCNP. The corresponding reactivity coefficient is thereby determined by dividing the reactivity change over the associated parameter changes. Core design studies are underway at NIST to develop a low-power LEU fueled beam reactor to advance the neutron source capability at NCNR for the next century. The reactivity effect caused by moderator temperature, fuel temperature and void is investigated though a direct perturbation approach using the MCNP code. The MTC, FTC, and VC for a representative startup core is obtained and the negative effect characteristics of these coefficients are observed. The reactivity effects under several postulated abnormal scenarios are also investigated. The overall negative feedback to the reactivity at each situation is confirmed. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Evaluation of VVER Reactivity Coefficients in Normal Operating Regimes with Change of External Load
The activity coefficients of a reactor facility with VVER in normal operating regimes associated with a change of the external load are studied. The model used to evaluate the reactivity coefficients in terms of the moderator density and fuel temperature aided by the connected KORSAR/GP code is described. A three-dimensional model of the neutron kinetics and a realistic model of the main regulators are used. Calculations of the reactivity coefficients for a regime with unloading of the turbogenerator and the regime of primary regulation of the power system frequency are studied as test problems. A physical analysis of the results is performed.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Reactivity Coefficient Measurements to Aid in Reducing Compensating Errors in Plutonium Nuclear Data
Compensating errors between several nuclear data observables in a nuclear data library can adversely impact application simulations. The primary goal of the EUCLID project (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) is to reduce compensating errors between fast (0.1–5 MeV) 239Pu nuclear data for prompt fission neutron spectra (PFNS), average prompt fission neutron multiplicities, and neutron induced fission, capture, elastic, and inelastic cross sections. This work will focus on the design and execution of void reactivity coefficient measurements in the EUCLID experiment, performed on the Planet vertical lift critical assembly machine at the National Criticality Experiments Research Center (NCERC). Two different base configurations were designed and measured, one with high neutron leakage, and one with low neutron leakage. Both were primarily made up of plutonium metal (Zero Power Physics Reactor plates) without interstitial moderators and reflected by half-inch aluminum. Design optimization showed that void reactivity coefficient measurements in three locations per configuration was most impactful to reduce nuclear data uncertainties due to the varying impacts from elastic and inelastic scattering, as well as fission and capture. The locations for measurements were chosen based on preliminary studies which balanced measurement uncertainty and measurement practicality. The measurements were also selected to have sensitivities maximally complementary to previous arrangements. Comparisons across nuclear data libraries highlight the potential impact.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Experiments to Measure the Inversion Point of the Isothermal Reactivity Coefficient in a Water-Moderated Pin-Fueled Critical Assembly at Sandia
A new set of critical experiments exploring the temperature-dependence of the reactivity in a critical assembly is described. In the experiments, the temperature of the critical assembly will be varied to determine the temperature that produces the highest reactivity in the assembly. This temperature is the inversion point of the isothermal reactivity coefficient of the assembly. An analysis of relevant configurations is presented. Existing measurements are described and an analysis of these experiments presented. The overall experimental approach is described as are the modifications to the critical assembly needed to perform the experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Experiments to Measure the Inversion Point of the Isothermal Reactivity Coefficient in a Water-Moderated Pin-Fueled Critical Assembly at Sandia [Slides]
Two experiment series are planned to measure temperature effects in the Sandia Critical Experiments. The first series (IER-304) will measure the critical size of a fuel rod configuration at several temperatures. The second series (IER-452) will measure the inversion temperature of the isothermal reactivity coefficient in the Sandia critical assembly. Each experiment in the second series will be preceded by one or more experiments in the first series. This presentation addresses the conceptual design of the inversion temperature experiments.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Reactivity Coefficient Measurements and Sensitivity Studies [Abstract]
Nuclear data validation is often performed today using criticality measurements. The gold standard for criticality measurements is the International Criticality Safety Benchmark Experiment Project (ICSBEP). The validation specifically focuses on the effective multiplication factor (k eff ). K eff is a relatively easy parameter to infer and has reduced uncertainty due to being at or above critical. However, while k eff is a well-documented parameter with detailed sensitivity and uncertainty analysis, it cannot be used as a standalone metric to determine inaccuracies in nuclear data (e.g., cross section data, PFNS, nu), which is based on theory, physics, and differential measurements. The Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) project aims to identify compensating errors in specific nuclear data by optimally designing experiments that are sensitive to a suite of measurement parameters beyond k eff . By identifying how each parameter's nuclear data sensitivity differs from others, experiments can be designed to constrain questionable nuclear data. One sensitivity that is of particular interest to this project includes reactivity coefficient sensitivities.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Measuring the Inversion Point of the Isothermal Reactivity Coefficient in a Water-Moderated Pin-Fueled Critical Assembly.
Abstract not provided.
Harms, Gary A.↗
Time Scale Effects on the Temperature Coefficient of Reactivity
For a nuclear reactor, the temperature coefficient of reactivity (TCR) is of significance for both power control and safety analysis, but it is not always a constant. For example, the TCR may be dependent upon temperature due to Doppler broadening, and may be also dependent upon time scale due to heat transfer and thermal expansion. A fast burst reactor is a favorable nuclear facility to simulate the time scale effects on the TCR. Fast burst reactors are generally unmoderated research reactors of enriched uranium metal or uranium alloys. The neutrons are fast and the Doppler broadening effects can be neglected, and the TCR is mainly contributed by fuel thermal expansion. Fast burst reactors can operate on both steady state power mode and burst power mode. In steady state power mode, the systems are assumed to be in thermal equilibrium. In burst mode, the full width at half maximum (FWHM) of the power is from tens of microseconds to a few milliseconds, and the heat transfer is generally neglected. If FWHM is short enough, thermal expansion will have a time delay due to inertial effect. This paper is aimed to study the exact time scale effects on the TCR due to thermal expansion. The TCR contributed by the Doppler broadening effects was neglected, because it is relatively small (The k{sub eff} as a function of temperature was also calculated using the Monte Carlo particle transport code JMCT by our group, and no definite difference due to the temperature change was obtained near room temperature), and there is no method to evaluate the time scale dependence of the Doppler broadening effects yet. In the time scale from 1 x 10{sup -6} s to 1 x 10{sup 5} s, the TCR of a Godiva-like spherical fast burst reactor was calculated with perturbation theory. A Godiva-like spherical fast burst reactor was studied, the calculation of the time scale effects on TCR involved thermal conduction, thermoelasticity and reactivity variation analysis. Both the thermal and the thermoelastic problems were solved using ANSYS software. ANSYS is software implementation of finite element simulations, which is widely used in engineering field. The reactivity variation due to thermal expansion was calculated with perturbation theory.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Assessment of Core Physics Characteristics of Extended Enrichment and Higher Burnup LWR Fuels using the Polaris/PARCS Two-Step Approach. Vol. I: PWR Fuel
Nuclear fuel with extended enrichment (235U enrichment within 5-8 wt%) is one of the evolutionary changes that have been pursued in recent years by commercial light water reactor operators and fuel vendors to improve the fuel cycle economy and operation performance of a nuclear plant. This work assesses the performance of the Polaris/PARCS two-step approach in core physics modeling of the pressurized water reactor cores with extended enrichment fuel, referred to as “LEU+” in this report. A representative LEU+ core with a 24-month fuel cycle developed by Southern Nuclear Company (SNC) was modeled using this two-step approach. A representative LEU core with an 18-month fuel cycle was also modeled to provide a reference for the LEU+ core. As expected, significantly more burnable poison absorbers were used in the LEU+ core to accomodate its higher fuel enrichment. Nine different fuel assembly types were modeled using Polaris for each core to generate the assembly cross sections, which were then processed by GenPMAX to prepare the cross-section data for PARCS. The average specific powers of each fuel batch in each core were derived from VERA results and higher specific powers in fresh assemblies were found in the LEU core due to its less total uranium loading included in the VERA LEU model, given that the total core power was assumed to be the same for both cores. PARCS models were developed to simulate the steady-state operations of both cores. PARCS results on the LEU+ core were first compared with the VERA results for verification purpose; good agreements were seen in soluble boron and burnup distribution results, indicating that the Polaris/PARCS modeling and simulation were correctly implemented. Core physics parameters calculated by PARCS, at zero power physics tests, beginning of cycle (BOC), and end of cycle conditions (EOC), were compared between the LEU+ core and the LEU core, including soluble boron concentration, burnup distributions, assembly and pin power peaking factors, fuel temperature reactivity coefficients, moderator temperature and density reactivity coefficients, control rod worth, and shut down margin. The main differences in PARCS results between the LEU+ and the LEU cores are summarized below: 1)The critical boron concentrations were found to be much higher in the LEU+ core than in the LEU core (1582 vs. 1335 ppm for peak values). 2)Higher assembly radial power peaking factors (1.4 vs. 1.3 for peak values), 2D pin peaking factors (1.53 vs. 1.42 for peak values), and 3D pin peaking factors (1.89 vs. 1.81 for peak values) were found in the LEU+ core than in the LEU core. 3)Significantly higher reactivity coefficients of moderator temperature (and density) were found in LEU+ than in LEU.4)Significantly lower control rod worth at EOC were found in LEU+ than LEU for all but one control banks.5)Significantly lower shut down margins were found in the LEU+ core than in the LEU core. The Polaris/GenPMAX/PARCS code suite was found to be capable of modeling the LEU+ PWR core for steady-state operations and no unexpected results in core physics parameters were observed, in spite of that a) several bugs in PARCS were identified and workarounds were used; b) several features were found lacking in the current version of PARCS that would be useful for core modeling. A list of requests for bug fixes and feature upgrades for PARCS originated from this work were transmitted to the code developers. The assessments on the performance of the Polaris/PARCS two-step approach in core modeling for boiling water reactor with LEU+ fuel is ongoing.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Sensitivity-based Similarity Metrics for New Experiment Design Optimization
The nuclear data used in advanced reactor simulations requires validation. Data from nuclear criticality experiments can provide this validation. New nuclear criticality experiment design requires extensive knowledge and expert judgement such that the experimental design parameters are selected in such a way to keep the experiment subcritical. To aide in this experimental design process, professionals can utilize sensitivity and uncertainty analysis. Sensitivity and uncertainty analysis relies on matching new application experiments with currently existing benchmark experiments. Currently, there is functionality in the Whisper 1.1 software package to calculate a similarity metric based on neutron multiplication factor sensitivity coefficients between a new application designed by the user and existing International Criticality Safety Benchmark Experiment Project (ICSBEP) benchmarks. The Whisper 1.1 software package is included in Monte Carlo N-Particle ® Code Version 6.21 (MCNP ® 6.2). This work is geared toward expanding this capability to new similarity metrics based on beta-effective sensitivity coefficients and reactivity coefficient sensitivity coefficients. While the investigation of these sensitivity coefficients is presented in detail in separate works at this same conference, this work will be primarily focused on studying the similarity metrics in more detail. These similarity metrics will then be incorporated into the optimization algorithms used for experiment design in EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data), which is a Los Alamos National Laboratory (LANL) project designed to constrain nuclear data of interest, such that adjustments can be made to possible inaccuracies. A more detailed optimization can be subsequently performed by breaking down these similarity metrics by isotope, reaction, and energy.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Simulation of the Fast Reactor Fuel Assembly Duct-Bowing Reactivity Effect Using Monte Carlo Neutron Transport and Finite Element Analysis
This paper discusses a new method of simulating the fuel assembly duct-bowing reactivity coefficient for EBR-II run 138B. Quantification of the fuel assembly duct-bowing reactivity effect in liquid metal–cooled fast reactors has been a persistent problem since they were first designed and operated. Simulation of the duct-bowing reactivity effect is difficult because the level of detail required to simulate the effect has exceeded most modeling capabilities. The new method outlined in this paper utilizes the finite element analysis code ANSYS to analyze the thermal and structural components. Here, the displacement of the fuel assembly duct due to thermal expansion and mechanical interaction was calculated by ANSYS using recorded EBR-II run 138B temperature and power boundary value data. The displacement values were incorporated into to a Monte Carlo model of EBR-II run 138B and keff was calculated. Multiple Monte Carlo calculations were performed with duct displacement values corresponding to different reactor temperatures. Using the calculated keff values associated with the different duct displacement results allowed calculation of the duct-bowing reactivity coefficient. The duct-bowing reactivity coefficient was calculated to be –14.5 × 10 –4 $/°C/ ± 4.4%.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Preliminary neutronic assessment of a novel fluoride-salt-cooled high-temperature reactor design
The Fluoride-Salt-Cooled High-Temperature Reactor (FHR) is a molten salt cooled, graphite moderated high temperature reactor. In this work, single-batch cycle length, block power distributions, and temperature-related reactivity effects were assessed using the SCALE code suite. Two types of parametric studies were performed. The first consisted of fixed temperature (isothermal) cases in which to determine the relationship between operation power and cycle length as well as pin power distributions. This second study calculates various isothermal reactivity coefficients based on heterogeneous temperature distributions. Reactivity coefficients with respect to the temperature of the fuel, coolant, and moderator were performed separately to determine their individual contributions. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Code Coverage Status of the ARC Code PERSENT
The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. PERSENT fulfills the role of generating reactivity coefficients for a given time point of a REBUS calculation usable in a point kinetics based safety analysis capability. PERSENT also provides a sensitivity coefficient capability on eigenvalue, reactivity worth, and several other key coefficients that are used in the follow-on safety analysis. Given a co-variance matrix, PERSENT can carry out the uncertainty quantification to indicate the amount of error in the reactivity coefficients derived from the errors in the cross section measurements. With continued improvement of computational resources, many of the geometry modeling capabilities in DIF3D that were primarily used in low order schemes are not really needed anymore. Today, the diffusion and transport capabilities of DIF3D-VARIANT are primarily used in the reactor design process with some scattered usage of DIF3D-FD and DIF3D-Nodal. PERSENT is part of the ARC code system and is built around DIF3D-VARIANT and the flux solution it provides. The purpose of the present work is to identify a set of test problems for PERSENT and assess the code coverage of PERSENT for those test problems. PERSENT treats the DIF3D executable as an external executable and thus the code coverage considerations only need to focus on the PERSENT source code and only a fraction of the connected modules in the existing ARC software library. The goal is to document what parts of the existing PERSENT code are touched by the set of test problems and which are not. Because the verification work done on PERSENT was focused on the most common uses of PERSENT for fast reactor analysis, the code coverage assessment of those capabilities is the highest priority. This will ensure that nothing is being missed by the existing verification test problems that users of PERSENT rely upon. The code coverage analysis of PERSENT was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of PERSENT. The detailed coverage tables are given for each submodule of PERSENT. Most of the uncovered parts/files could be easily ignored because they are either for error message and debugging output or not needed by PERSENT today. Only a few uncovered parts of PERSENT deserve extending the verification test suite.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Validation of MCNP5 for use in Calculating Temperature Coefficients of Reactivity for the SHINE System
Neutron transport calculations for the SHINE system are being done with MCNP, a general-purpose Monte Carlo NParticle code developed at Los Alamos National Laboratory (LANL). Current calculations are being done with MCNP5, version 1.6. This paper describes the methods used to calculate a tolerance band for temperature coefficients of reactivity (α{sub T}) calculated with MCNP5, v1.6 for the SHINE system. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗