Evaluated Criticality Safety Benchmark Experiments Research Articles

Analysis of Major Benchmark Uncertainties for Fast Metal Assemblies in the ICSBEP Handbook

The International Handbook of Evaluated Criticality Safety Benchmark Experiments includes thousands of criticality safety benchmark configurations that have been collected since the early 1990s. Their quality and uncertainty analysis have greatly expanded over the years, as computer codes have also expanded and accelerated. Along with the increase in effort required for benchmarks today, the cost of performing experiments has also gone up. This paper evaluates plutonium fast metal and highly enriched uranium fast metal experiments for their major uncertainty contributions and identifies where to focus efforts in future experiment design.

Comparison of two methods for assessment of the rod positioning uncertainty and consequences on the evaluation of correlation factors

In this paper two major families of methods to deal with the assessment of the rod positioning uncertainty in a lattice are tested: a traditional one described in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) Handbook and the other one consisting in sampling the position of rods with Monte Carlo techniques (ISO Uncertainty Guidelines). They are applied on a benchmark with tight-packed lattice of UO2 rods that is sensitive to the rod positioning as it is clearly under-moderated. It is shown that the choice of the method has a great impact on the propagated uncertainty, the traditional one leading to a significant overestimation of the overall uncertainty and can also contribute to a bias in the correlation factors that are used for assessing biases due to nuclear data using GLLSM methodologies. The paper briefly describes the tight-packed lattice experimental program performed at the Valduc Research Centre, which is at the origin of these concerns. Then it proposes a simple model on which to apply simulations of rod positioning to be performed with MORET 5 Monte Carlo code using the Prométhée tool. Results demonstrate that use of Monte Carlo methodologies provide more realistic uncertainty estimates in fuel pitch that are consistent with repeatability/reproducibility experiments. The current comparisons use light water reactor systems, which is directly relevant to some small modular reactor designs. However, accurate prediction and estimate of uncertainties in pitch for advanced reactor systems is also relevant. The application of unrealistic uncertainty analysis methods can incur larger margins in advanced reactor design, safety, and operation than are necessary.

On a comparison of a neutron Monte Carlo transport simulation to a criticality benchmark experiment

In this work the Water-Reflected 91-Liter Sphere of Enriched Uranium Oxyfluoride Solution experiment from the International Handbook of Evaluated Criticality Safety Benchmark Experiments is simulated with a neutron transport Monte Carlo simulator. The simulator works with continuous variables of the phase space relevant in transport theory, which allows to extract the simulated spectral neutron flux and render it in parameterized representation. The neutron effective multiplication factor is calculated and compared to the experimental findings, as well as to results from other simulation codes. Agreement between experiment and the present simulation was found to be within one standard deviation.

An inter-comparison between ENDF/B-VIII.0-NECP-Atlas and ENDF/B-VIII.0-NJOY results for criticality safety benchmarks and benchmarks on the reactivity temperature coefficient

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutual-comparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

Application of deterministic sampling methods for uncertainty quantification in manufacturing tolerances in neutron physics

In this work, an application of the deterministic sampling (DS) methodology for uncertainty quantification of manufacturing tolerances is assessed. The DS methodology is applied to eight benchmark cases, from two criticality safety benchmark series, extracted from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (i.e., LCT-007 & LCT-039). The DS results are compared to two other sampling techniques: Random Sampling and Latin Hypercube Sampling. Ten input parameters such as isotopic composition of the fuel, geometry properties of the assembly, pitch array, etc. were considered. Only the multiplicative factor (keff) was considered as output quantity. Evaluation of the keff was obtained using the Monte Carlo code MCNP for the LCT-007 case, and the Serpent code for LCT-039.Results show very good agreement among all methodologies for the calculated keff values and their corresponding uncertainties. The largest uncertainty difference with DS, compared to the reference values, is 46 pcm for LCT-007 and 9 pcm for LCT-039. The use of DS remarkably reduces the number of samples needed to perform an accurate uncertainty quantification (from thousands of samples down to a dozen).This study provides insights in the application of DS methodology in the neutron physics domain and showcases its benefits and limitations.

DIAGNOSIS OF THE UNRESOLVED DOMAIN TREATMENT IN MONTE CARLO TRANSPORT CALCULATIONS THROUGH THE IDENTIFICATION AND MODELLING OF CRITICALITY SAFETY EXPERIMENTS

Monte Carlo neutron transport codes can be used for high-fidelity predictions of the performance of nuclear systems. However, validation against experiments is required in order to establish the credibility in the results and identify the inaccuracies due to the used calculation scheme and associated databases. The International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) contains criticality safety benchmarks derived from experiments that have been performed at various nuclear critical facilities around the world and are very valuable for validation purposes. The main objective of this work is the identification and modelling of experimental benchmarks included at ICSBEP in support of the validation of Monte Carlo neutron transport calculations when applied to fast systems, and in particular, KENO-VI and associated AMPX-formatted continuous-energy libraries from SCALE package. In such systems, the predicted k-eff values can be very sensitive to the treatment of nuclear data in the Unresolved Resonance Region (URR). Consequently, benchmarks with intermediate and fast spectra are identified and modelled with KENO-VI. Then, calculated results with and without probability tables in the URR are compared with each other in order to identify the most sensitive configurations to the URR. As a result of the proposed study, recommendations are given about the benchmarks that should be modelled and analysed to qualify the processed continuous-energy libraries before their use in Monte Carlo transport codes for practical fast reactor applications.

NEUTRONIC BENCHMARKING OF SMALL GAS-COOLED SYSTEMS

To demonstrate that nuclear reactors can be built faster and more economically than they have been in the past, the US Department of Energy Office of Nuclear Energy is sponsoring the development of a small nuclear reactor called the Transformational Challenge Reactor (TCR) [1–2]. An important part of the design and licencing process of a new reactor is validation of the software used to analyze the reactor using established reactor physics benchmarks. This paper discusses validation of the neutronics software used to model four preliminary designs of the TCR core [2]. Because the TCR core design uses innovative technology and methods, comparable established benchmarks are limited or do not exist. For this effort, established benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) [3] were considered to be suitable for this design based on analysis using the SCALE/TSUNAMI-computed similarity indices to determine the amount of shared uncertainty between the design and each selected benchmark experiment. This paper addresses the challenges faced in benchmarking a unique reactor for licensing and construction, a task that will become more common as a new generation of innovative nuclear reactors are designed and built.

Trend in criticality calculation results of borated-water-moderation experimental cores relating to 10B(n, α) cross section

ABSTRACT In connection with the accuracy of the 10B(n, α) cross section in the thermal- and epithermal-neutron energy regions, criticality calculation results were examined for six benchmark sets of light-water-moderation critical experiments of UO2 and MOX fuel lattice cores with un-borated and borated water. Two of the benchmark sets were those implemented in the Tank-Type Critical Assembly (TCA). The others were taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP), and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). The enrichments of the UO2 fuel range from 1.9 wt% to 2.6 wt%, and the Pu contents of the MOX fuel do from 2.0 to 6.6 wt%. The boron concentrations in water are up to 1511 ppm. The effective neutron multiplication factors (keff ) were taken from the published documents. They were calculated with continuous-energy Monte Carlo calculation codes in combination with JENDL-4.0, and other evaluated nuclear data libraries. It was confirmed that the keff values of the critical cores increased with the boron concentrations, which indicates that the 10B(n, α) cross section in the thermal- and epithermal-neutron energy regions should be larger than those in JENDL-4.0 and other libraries.

Resonance parameter adjustment in the resolved region based upon an Asymptotic Generalized Linear Least-Squares methodology in conjunction with the Monte Carlo method

In order to reduce methodological effects, a stochastic technique is proposed for adjusting basic ENDF parameters instead of preprocessed multi-group data as it is normally the case; the required sensitivity coefficients of the integral parameters to these parameters are recomputed by means of NJOY/Serpent within an iterative scheme based upon the Asymptotic Progressing Incremental nuclear data Adjustment (APIA) methodology. It is anticipated that the provided illustrative examples including selected benchmarks along with ENDF parameters to adjust are certainly limited, making the study too preliminary in order to reach meaningful conclusions on the quality of the posterior nuclear data. However, they were conceived for showing how the proposed novel methodology along with its main principles could be generalized and equally used in the framework of a future exhaustive adjustment, largely justifying this paper.More specifically dealt with in this study is the assimilation of the effective multiplication factor for low-enriched, U-solution, thermal systems by adjusting resonance widths for U-235 and U-238. It is shown that for the experiments part of the assimilation, the posterior ratios of computed to experimental central values are exactly equal to one, reflecting the general situation. This outcome may be expected from the theory, provided that convergence can be reached, which is always the case if the chosen experiments are not too strongly correlated among each other from the analytical viewpoint. Otherwise, differently from non-iterative methods, the procedure, since diverging, cannot provide useful data: in this case, some experiments of less relevance should be withdrawn.A similar approach could thus be used by considering a wide database in conjunction with a broader selection of basic parameters, to provide adjusted ENDF format files ensuring, by means of a reference stochastic tool, equality between computed and experimental values for a large variety of experimental data e.g. those available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP).

Contributions to integral nuclear data in ICSBEP and IRPhEP since ND2016

The contributions to the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) was last presented to the international nuclear data community at ND2016. Since ND2016, integral benchmark data that are available for nuclear data testing has continued to increase. The 2018 edition of the International Handbook of Evaluated Criti-cality Safety Benchmark Experiments (ICSBEP Handbook) now contains 574 evaluations with benchmark specifications for 4,916 critical, near-critical, or subcritical configurations, 45 criticality alarm placement/shielding configuration with multiple dose points apiece, and 215 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. The 2018 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) contains data from 159 different experimental series that were performed at 54 different nuclear facilities. Currently 156 of the 159 evaluations are published as approved benchmarks with the remaining three evaluations published as drafts. Measurements found in the IRPhEP Handbook include criticality, buckling and extrapolation length, spectral characteristics, reactivity effects, reactivity coefficients, kinetics, reaction-rate distributions, power distributions, isotopic compositions, and/or other miscellaneous types of measurements for various types of reactor systems. Additional benchmark evaluations will be included in the 2019 editions of these handbooks. These handbooks continue to represent the standard for neutronics benchmark experiment evaluation.

Analysis of the Recent Nuclear Data through a Fast Benchmarks Highly Enriched Uranium Study Using the Mcnp6.1 Code

In this work, we present our MCNP6.1 modeling of some critical fast experimental benchmarks, aiming to qualify our cross section libraries derived from ENDF/B-VII.1, ENDF/B-VII, JEFF-3.1, JENDL-3.3, and JENDL-4.0. The analyzed benchmarks are characterized by simple geometries which helps to have taken precise results, and concerning the type: HEU-MET-FAST (highly enriched uranium). Those benchmarks are extracted from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) published by the Nuclear Energy Agency [1]. A detailed comparison of the results of our simulation was made in order to highlight the influence of these nuclear data types on our calculations, due to its importance for the stability of nuclear reactors. We interpreted the difference between calculation and experiment ($$C{-}E$$) for the principal parameter $$k_{\textrm{eff}}$$ through a comparison of the fission and capture rates of the major fissile elements. The different spectral indices F28/F25, F49/F25, F37/F25, C28/F25, and F23/F25 at the cores center are also calculated. For the majority of the studied HEU (highly enriched uranium) benchmark cases, The ENDF/B-VII and JEFF-3.1 have a good agreement with the experimental ones, concerning the $$k_{\textrm{eff}}$$ results. The average discrepancy from the experimental values for ENDF/B-VII is 0.42$$\%$$, and 0.39$$\%$$ for JEFF-3.1. An overestimation was observed for most evaluations concerning benchmarks with tungsten carbonate reflectors. The best results were obtained by JENDL-3.3, with a maximal discrepancy $$C{-}E$$ estimated 0.47$$\%$$ concerning fission rate, and 4.25$$\%$$ for capture rate. In analyzing the spectral indices, for GODIVA and FLATTOP-25, the best results were obtained by JENDL-4.0 with a maximal discrepancy of 2.66$$\%$$.

235U elastic cross-section adjustment in criticality benchmarks – Comparison between JENDL-4.0 and ENDF/-VII.1

The main aim of this study is to estimate the uncertainty on the multiplication factor (keff) caused by the elastic cross-section uncertainty of the 235U nucleus, and the adjustment of this cross-section using the generalized linear-least squares method. For this purpose, the sensitivity matrix of keff with respect to 235U elastic cross section in the two nuclear evaluations ENDF/B-VII.1 and JENDL-4.0, has been assessed in several thermal, intermediate and fast benchmarks; which have been taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). The selection of these benchmarks is based on chi-square statistic test (χ2) that measures the statistical weight of deviations between calculus and experience results of the integral parameters caused by uncertainty in differential data. In this statistical test the sensitivity and covariance matrices are used. The sensitivity matrix has been calculated by the adjoint-weighted perturbation technique of the multi-purpose Monte Carlo code MCNP6.1. The 235U elastic covariance matrix has been generated in 44 neutron energy groups by the ERRORJ module of the nuclear data processing system NJOY99. The prior and posterior uncertainties on the keff, the adjusted elastic cross-section of the 235U and covariance matrix have been estimated using FORTRAN 95 code developed in our laboratory based on the generalized linear-least squares method.The results obtained show that: the 235U elastic cross-sections taken from JENDL-4.0 requires an adjustment which can reach 6% in 0–100 keV, 4.5% in 100 keV–625 keV and 2.5% in 625 keV–0.5 MeV, while those from ENDF/B-VII.1 needs an adjustment which can reach 5% in 0–100 keV, 3% in 100 keV-625 keV and 1% in 625 keV–0.5 MeV interval. The adjusted cross-section in the two evaluations ameliorates the posterior calculated keff with respect to the experimental keff. Also the posterior covariance matrix decreases the nuclear uncertainty of the adjusted keff.

Criticality validation of SuperMC with ICSBEP

Monte Carlo transport codes are extensively used in neutronic analysis, especially in criticality safety analysis and shielding analysis. Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC) is a CAD based Monte Carlo program for integrated simulation of nuclear systems. The aim of this paper was to show the capability of SuperMC on criticality calculation with different models. In this study 119 representative benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) were used for validating the code. All the benchmark models and input files were built and completed by employing the models’ conversion and construction function of SuperMC. The results showed that there was a good agreement between SuperMC and experimental data, and the discrepancies were mainly included in the statistical uncertainty.

Criticality safety analysis using continuous energy libraries of MCNP code

The study of subcritical and critical systems is useful to guide decisions about design, installation and operation of some devices, mainly nuclear reactors and power plants. The information generated by these systems guides the best decisions to be made in executive project, economic viability and the safety measures to be employed in a nuclear facility. Simulating some experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments, the MCNP5 code has been validated as an important tool to make an analysis about nuclear criticality safety. Its continuous libraries have been used. The average values and the standard deviation (SD) are evaluated. The results are very close to the values obtained by the benchmark experiments and in agreement with the criticality safety limits required by the Brazilian authorities.

The MIRTE Experimental Program: An Opportunity to Test Structural Materials in Various Configurations in Thermal Energy Spectrum

The MIRTE (Materials in Interacting and Reflecting configurations, all Thicknesses) program was established to answer the needs of criticality safety practitioners in terms of experimental validation of structural materials and to possibly contribute to nuclear data improvement, which ultimately supports reactor safety analysis as well. MIRTE took the shape of a collaboration between the AREVA and ANDRA French industrialists and a noncommercial international funding partner such as the U.S. Department of Energy. The aim of this paper is to present the configurations of the MIRTE 1 and MIRTE 2 programs and to highlight the results of the titanium experiments recently published in the International Handbook of Evaluated Criticality Safety Benchmark Experiments.

Criticality Benchmarks for the Epithermal Test Assembly Experiments

The Epithermal Test Assembly (ETA) experiments were performed to test the adequacy of 233U, 235U, and 232Th cross sections in epithermal spectra in support of the Light Water Breeder Reactor (LWBR) Program. The ETA design contained a central heavy water–moderated test region surrounded by a light water–moderated annular driver region. Two series of experiments were performed: ETA-I with 235UO2-ThO2 fuel rods in the test region and ETA-II with 233UO2-ThO2 fuel rods in the test region. The dominant uncertainties in the critical configurations include the test-rod pitch, fuel density, and ThO2 mass for ETA-I; the test-region fuel-rod fuel density and 233U to (233U + Th) weight ratio for ETA-II; and the driver-region fuel-rod outer diameter, uranium enrichment, and pitch for both ETA experiments. Benchmark model results using MCNP5 are provided for ENDF/B-V, ENDF/B-VI, ENDF/B-VII.0, and ENDF/B-VII.1 cross sections with only the ENDF/B-VII.0 results falling within three standard deviations of the benchmark model keff. The ETA-I and ETA-II benchmark evaluations have been included in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and are replicated in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

Evaluation of Cadmium Ratio and Foil Activation Measurements for a Beryllium-Reflected Assembly of U(93.15)O2 Fuel Rods (1.506-cm Triangular Pitch)

A series of small, compact critical assembly experiments was completed from 1962 to 1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments program. Initial experiments, performed in November and December 1962, consisted of a core of unmoderated stainless steel tubes surrounded by a graphite reflector. Later experiments included beryllium-reflected assemblies with the fuel in a 1.506-cm triangular lattice and in seven-tube clusters. Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, and cadmium ratios were performed. The critical configurations, the cadmium ratio, and activation rate measurements for the beryllium-reflected 1.506-cm-array critical configuration have been evaluated and are described in this paper. It was found that these measurements are acceptable as benchmark experiments and have been included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments and the International Handbook of Evaluated Criticality Safety Benchmark Experiments.

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (keff) Predictions

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. This paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methods and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth within the application model. This paper provides a detailed description of the approach and its technical bases, describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models, and provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data.

An Overview of the Experiments Performed at the BFS Facilities and Evaluated for the International Reactor Physics Experiment Evaluation Project

In total, eight evaluations with 27 critical configurations spanning 11 series of assemblies conducted at the BFS-1 and BFS-2 facilities [BFS: Bol’shoy Fizicheskiy Stand (in transcription from Russian: Big Physical Facility)] are documented in the 2013 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). The evaluations concern four scientific issues: physics of sodium-cooled, medium-sized fast reactors; physics of sodium-cooled, small-sized experimental reactors; physics of lead-cooled fast reactors; and integral benchmark data for the mixed oxide (MOX) fuel manufacturing process, particularly low-moderated MOX fissile media. Also, two evaluations with 11 critical configurations spanning two series of assemblies conducted at the BFS-1 facility where many physical parameters in addition to criticality were measured and are documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). These experiments were designed to obtain data that are applicable to a wide range of criticality safety operations involving fissile contaminated waste streams. This paper provides an overview of the BFS experiments presented in the IRPhEP and ICSBEP Handbooks.

Establishment of Correlations for Some Critical and Reactor Physics Experiments

Criticality safety and reactor physics benchmark experiments may share certain components of the configurations and the same experimental techniques. This results in correlations between pairs of benchmark experiments. The correlations may impact validation results of criticality safety or reactor physics calculations as well as nuclear data validation. This paper presents how experimental correlations have been established for some fast and thermal neutron spectra configurations, documented in International Handbook of Evaluated Criticality Safety Benchmark Experiments and in International Handbook of Evaluated Reactor Physics Benchmark Experiments. It also discusses the impact of those correlations on validation of the most recent nuclear data libraries.