Criticality Safety Evaluations Research Articles

Addressing practical aspects of the nuclear criticality safety evaluations with frequentist statistics

General guidelines for Criticality Safety Evaluations (CSE) and regulatory requirements on the Nuclear Criticality Safety are well established nowadays. However, the expert knowledge is often still required in practice. Such situation is typical for the CSE methodologies grounded on the frequentist statistic inference and involving tolerance limits estimations and hypothesis testing. In this respect, various sensitivity tests are illustrated in the given paper to enhance understanding of the statistical assessments, in application to conventional CSE for Light Water Reactor (LWR) fuel configurations, when relatively small validation samples are generally available. It is further discussed that for rare situations with very small validation data samples, both parametric and non-parametric statistical methods have drawbacks: unreliability of normality tests for the first case and large non-parametric margins historically recommended for the second case. For such specific situations, an additional safety margin named “Normality Margin” is proposed in this work, towards compensation of the lack of knowledge of the underlying hypothetical probability distribution describing the criticality validation results.

Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

A High-Assay Low-Enriched Uranium Fuel Transportation Concept

The uranium 235U enrichment commonly used in fuel production for U.S. light water nuclear reactors typically does not exceed 5 wt%. In contrast, many of the currently investigated advanced reactor concepts demand fuel with higher enrichments. This includes high-assay low-enriched uranium (HALEU), characterized by a 235U enrichment of 5 to 20 wt%. The necessity of HALEU transportation in the fuel production cycle leads to new challenges caused by various technical and regulatory hurdles. Current U.S. Nuclear Regulatory Commission–approved transportation package designs for UF6 with enrichments above 5 wt% provide relatively small payloads [≤116 kg (250 lb)]. Furthermore, in accordance with 10 CFR 71.55, package design activities for fissile material enriched above 5 wt% need to consider water infiltration in the containment as part of the criticality safety evaluations. This study presents a transportation package concept for HALEU advanced nuclear reactor fuel with a significantly higher payload of up to 376 kg (830 lb) of fissile material per package and up to 1881 kg (4149 lb) of HALEU per legal weight truck. The anticipated chemical form of the transported material is UO2 downblended from available highly enriched uranium. The concept utilizes a combination of existing transportation packaging, 18 inner canisters, and a novel basket design that includes a borated aluminum flux trap. Criticality and shielding evaluations; fundamental structural, confinement, and thermal assessments; and studies on package operations are presented. The results of this study build significant confidence in the technical feasibility of a high-capacity HALEU transportation package concept while demonstrating the concept’s potential to meet U.S. regulatory requirements.

On the options for incorporating nuclear data uncertainties in criticality safety assessments for LWR fuel

In view of massive operations with LWR spent nuclear fuel for the forthcoming final geological disposal, as expected, e.g., in Switzerland, needs are growing for exhaustive and enhanced criticality safety assessments to allow optimization of the spent nuclear fuel transportation, storage and disposal strategies. At the Paul Scherrer Institute (PSI) the current base methodology for criticality safety evaluations (CSE) used for Swiss applications is focused, following common practices, on establishing an upper subcritical limit (USL) for the effective neutron multiplication factor (keff), based on a statistical evaluation of the obtained validation results in terms of the calculated-to-benchmark keff values. Recently, new calculation capabilities have been developed at PSI in particular to propagate neutron-nuclear data (ND) uncertainties in calculations with Monte Carlo transport codes in combination with general-purpose ND libraries. The computation tool NUSS for random sampling the ACE formatted ND libraries was developed for that purpose. Thereby, this paper concerns how these new calculation capacities could improve the present PSI CSE methodology and the prospects for its upgrading towards burnup credit applications are consequently discussed.

Review calculations for the OECD/NEA Burn-up Credit Criticality Safety Benchmark

A calculation methodology for criticality safety evaluations related to medium term dry storage, e.g. interim storage, and long term waste disposal, e.g. deep geological repository, is under development at PSI.In that sense, the performance of present decay and criticality safety codes has to be evaluated for this kind of applications, with code to code, and data to data, comparison as a very first step. In this article, the Burn-up Credit Criticality Safety Benchmark Phase VII organized by OECD/NEA has been employed to evaluate decay calculations results for the codes CINDER and SERPENT2. The computed isotopic compositions have then been used for the assessment of the criticality calculations with MCNP6 and SERPENT2, while employing the most recent available cross sections libraries from ENDF/B-VII.1, JEFF-3.2 and TENDL-2014.Results show overall a good agreement for all the options listed, while differences are pointed out and their origin is discussed.

Advanced calculation methodology for manufacturing and technological parameters’ uncertainties propagation at arbitrary level of lattice elements grouping

A new methodology, referred to as manufacturing and technological parameters uncertainty quantification (MTUQ), is under development at Paul Scherrer Institut (PSI). Based on uncertainty and global sensitivity analysis methods, MTUQ aims at advancing state-of-the-art for the treatment of geometrical/material uncertainties in light water reactor computations, using the MCNPX Monte Carlo neutron transport code. The development is currently focused primarily on criticality safety evaluations (CSE). In that context, the key components are a dedicated modular interface with the MCNPX code and a user-friendly interface to model functional relationship between system variables. A unique feature is an automatic capability to parameterize variables belonging to so-called “repeated structures” such as to allow for perturbations of each individual element of a given system modelled with MCNPX. Concerning the statistical analysis capabilities, these are currently implemented through an interface with the ROOT platform to handle the random sampling design. This paper presents the current status of the MTUQ methodology development and a first assessment of an ongoing organisation for economic cooperation and development/nuclear energy agency benchmark dedicated to uncertainty analyses for CSE. The presented results illustrate the overall capabilities of MTUQ and underline its relevance in predicting more realistic results compared to a methodology previously applied at PSI for this particular benchmark.

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.

Criticality safety analysis for a core catcher designed in Korea

Corium is a molten mixture of portions of a reactor core generated by a core melting accident. Corium includes fissionable materials; therefore, a criticality safety analysis must be performed for the core catcher design. This study analyzes the criticality safety of corium arranged in a core catcher developed in Korea. The corium composition was calculated for a 1400MWe nuclear power plant. There are several variables involved in the criticality evaluation of corium, thus conservative assumptions were used to reduce the number of variables. A criticality evaluation procedure was employed to assess the operational failure of the core catcher under different accident scenarios. Four kinds of scenarios were selected, and criticality evaluations were pursued for each case. The multiplication factors in each condition were calculated with MCNP5 code. Also, the code bias was calculated with the benchmark problems of 262 LEU experiments to account for the uncertainty of MCNP code. All evaluation results for the assumed scenarios showed that the core catcher satisfies the regulatory guidelines for criticality safety. The calculation results will be used in the design of a core catcher being developed in Korea. It is expected that the data calculated in this study can be used as reference data for criticality safety evaluations of core melting accidents. Also, the procedure for criticality safety evaluation proposed in this study can be utilized to establish regulatory guidelines in Korea.

SCALE 6: Comprehensive Nuclear Safety Analysis Code System

Version 6 of the Standardized Computer Analyses for Licensing Evaluation (SCALE) computer software system developed at Oak Ridge National Laboratory, released in February 2009, contains significant new capabilities and data for nuclear safety analysis and marks an important update for this software package, which is used worldwide. This paper highlights the capabilities of the SCALE system, including continuous-energy flux calculations for processing multigroup problem-dependent cross sections, ENDF/B-VII continuous-energy and multigroup nuclear cross-section data, continuous-energy Monte Carlo criticality safety calculations, Monte Carlo radiation shielding analyses with automated three-dimensional variance reduction techniques, one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations, two- and three-dimensional lattice physics depletion analyses, fast and accurate source terms and decay heat calculations, automated burnup credit analyses with loading curve search, and integrated three-dimensional criticality accident alarm system analyses using coupled Monte Carlo criticality and shielding calculations.

The effect of modern thermal neutron scattering sublibraries on criticality safety evaluations of wet storage pools

Within the framework of a recent criticality safety evaluation (CSE) performed for the licensing of a new commercial wet storage pool using MCNPX-2.5.0 along with the ENDF/B-VII.0 and JEFF-3.1 continuous energy cross-section libraries, the maximum permissible initial fuel-enrichment limit for water reflected configurations was found to be dependant upon the applied neutron cross section library. More detailed investigations showed that the difference is mainly caused by different sublibraries for thermal neutron scattering based on parameterizations of the S ( α , β ) scattering matrix. This paper addresses the main findings of these investigations and emphasizes which configurations are particularly sensitive to the parameterization of S ( α , β ) applied in the CSE analysis. The results are presented in an attempt, on the one hand, to assess more precisely the sensitivity upon thermal scattering modeling data in CSE of wet storage pools, and on the other hand, to assess whether the data employed in the most recent neutron data libraries can be considered as an improvement for this type of analyses.

Towards the development of upper subcriticality limits on the basis of benchmark criticality calculations

This paper concerns the assessment of standard point-wise neutron data libraries for criticality safety evaluations in units of the effective neutron multiplication factor, keff, the aim being to establish a methodology for the analysis of storage pools containing fuel assemblies discharged from the Swiss Light Water Reactors. The selected approach is based on using the Monte Carlo code MCNPX (version 2.4.0 was applied in the study at hand) and a modern standard point-wise neutron data library officially distributed by OECD/NEA databank. The approach is oriented towards meeting the broadly accepted general requirements to establish subcriticality, such as those formulated in the ANSI/ANS-8.1-1998 and ANSI/ANS-8.17-2004 Standards.In the above perspective, the results of the assessment of the standard neutron data libraries JEF-2.2 and JENDL-3.3 for criticality safety analysis of UO2 light water reactor fuel assemblies immersed in water, are provided and discussed. The assessment has been performed on the basis of a suite of low-enriched uranium thermal compound benchmarks selected from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Special emphasis is given to the appraisal of the applicability of the Gaussian distribution in the approximations and posterior analyses of the calculated benchmark results that are necessary to establish the keff safety margins. Such application has been found to be justified, on the one hand, by the observation of a very close agreement between parametric and non-parametric evaluations of the analyzed keffcalc/keffbench samples; on the other hand, one indeed expects the keffcalc/keffbench values of an ideal (cluster-less) set of criticality benchmarks with similar physical properties to follow a normal distribution.Nevertheless, we presume that similar deficiencies in the specification of the configurations belonging to one or similar series of experiments could cause unspecified experiment/evaluation-related systematic errors in the benchmark evaluations and consequently in the keffbench-values. Therefore, even distribution-free estimates will not be rigorous because they also rely on the assumption of random sampling of the keffcalc/keffbench population. Moreover, we suggest that the question of normality appears to be of minor importance compared to other hypotheses and to the conservative approximations typically assumed for criticality safety evaluations.

Effect of Neutron Induced Reactions of Neodymium-147 and 148 on Burnup Evaluation

Burnup is an important parameter in criticality safety evaluations of spent nuclear fuel in which burnup credit is taken into account. The Neodymium-148 method is widely used to evaluate the burnup of post irradiation examination (PIE) samples, and it is well known for its good accuracy. However, accuracy of the evaluated burnup values may be affected by the neutron capture reaction of 147Nd and 148Nd. Moreover, in the analysis of PIE data from a PWR, the calculation results of 148Nd have more than a 1% deviation from the experiment. In this study, the contribution of neutron capture reactions of 147Nd and 148Nd to the amount of 148Nd is discussed. The PIE data analyses using new evaluation of 147Nd capture cross section show that the JENDL-3.2 cross section data is overestimated. The change in the amount of 148Nd due to both reactions is less than 0.7% under normal reactor operation conditions. In particular, it is in the 0.1% range if burnup is approximately 30 GWd/t for a BWR and 40 GWd/t for a PWR.

Effect of Nuclear Fuel Particle Movement on Nuclear Criticality in a Rotating Cylindrical Vessel

In conventional criticality safety evaluations of nuclear facilities, nuclear fuel particle behavior could not be considered because any appropriate computational tools did not be developed. A new criticality evaluation tool, which can consider the particle behavior, is developed by coupling a Discrete Element Method with a continuous energy Monte Carlo type code. Criticality evaluations considering the nuclear particle movement in a rotating cylindrical tank are performed by applying this tool. Effects of the powder-filling ratio and the rotating speed on nuclear criticality are investigated in the present study. The nuclear powder free surface is fluctuated more remarkably as the filling ratio and the rotating speed are higher. The fluctuation makes the effective multiplication factor lower because of neutron leakage from the system by the increase of the superficial area and reduction of the atomic number densities due to the flowability.