Evaluated Criticality Safety Benchmark Experiments Research Articles

Evaluation of Delayed Critical ORNL Unreflected HEU Metal Sphere (ORSphere)

In 1971 and 1972 experimenters at the Oak Ridge Critical Experiment Facility performed critical experiments using an unreflected metal sphere of highly enriched uranium (HEU). The sphere used for the criticality experiments, originally used for neutron leakage spectrum measurements by General Atomic Company, consisted of three main parts and were assembled with a vertical assembly machine. Two configurations were tested. The first was nearly spherical with a nominal radius of 3.467 inches and had a reactivity of 68.1 ± 2.0 cents. The sphere parts were then re-machined as a sphere with a nominal radius of 3.4425 inches. This assembly had a reactivity of −23 cents. The method, dimensions, and uncertainty of the critical experiment were extensively recorded and documented. The original purpose of the experiments was for comparison to GODIVA I experiments. The ORNL unreflected HEU Metal Spheres have been evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (scheduled for inclusion in the September 2013 edition).

Identification of Possible Discrepancies in HEU Metal Benchmarks

The experiment from which benchmark specifications of “Lady Godiva” were derived consisted of nested hemispheres forming a bare sphere of highly enriched uranium (HEU). That experiment was performed in the early 1950s and the critical configuration was later (1995) evaluated and published in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The benchmark eigenvalue is reported as 1.000 ± 0.001, which is representative of a high-quality benchmark experiment. Our current neutronic codes and cross section data are tailored to provide qualitative results that concur with the GODIVA I benchmark. Since 1995, additional high-fidelity HEU metal benchmark data have been evaluated and published. Calculated results from many of those benchmarks are consistently low. While there is generally good agreement between calculated and benchmark eigenvalues for spherical and slab geometries, cylindrical configurations tend to calculate low.

Evaluation of Plutonium Hemisphere Critical Experiments Partially Reflected by Steel and Oil

A series of 15 critical experiments performed at the Rocky Flats Critical Mass Laboratory in the late 1960s was evaluated and then determined to represent acceptable benchmark experiments for the validation of calculational methods. This series of experiments was part of a larger set of experiments performed to evaluate operational safety margins at the Rocky Flats Plant. The experiments consisted of bare plutonium metal hemishells reflected by steel hemishells of increasing thickness and motor oil. The hemishell assembly was suspended within dual aluminum tanks. Criticality was achieved by pumping oil into the tanks such that effectively infinite reflection was achieved in all directions except directly above the assembly; then the critical oil height was recorded. The results of these experiments had been initially ignored because early computational methods had been inadequate to analyze partially reflected configurations. The dominant uncertainties include the uncertainty in the average plutonium density and the composition of materials in the gaps between the plutonium hemishells. Simple and detailed benchmark models were developed. Eigenvalue calculations using MCNP5 and ENDF/B-VII.0 were within 2σ of the benchmark values. This benchmark evaluation has been added to the International Handbook of Evaluated Criticality Safety Benchmark Experiments.

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

The ENDF/B-VII.1 library is the latest revision to the United Statesʼ Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., “ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,” Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Projectʼs International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U, 238,242Pu and 241,243Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for 233U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

Nuclear Data Performance Testing Using Sensitive, but Less Frequently Used ICSBEP Benchmarks

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) has published the International Handbook of Evaluated Criticality Safety Benchmark Experiments annually since 1995. The Handbook now spans over 51,000 pages with benchmark specifications for 4,283 critical, near critical, or subcritical configurations; 24 criticality alarm placement/shielding configurations with multiple dose points for each; and 200 configurations that have been categorized as fundamental physics measurements relevant to criticality safety applications. Benchmark data in the ICSBEP Handbook were originally intended for validation of criticality safety methods and data; however, the benchmark specifications are now used extensively for nuclear data testing. There are several, less frequently used benchmarks within the Handbook that are very sensitive to thorium and certain key structural and moderating materials. Calculated results for many of those benchmarks using modern nuclear data libraries suggest there is still room for improvement. These and other highly sensitive, but rarely quoted benchmarks are highlighted and data testing results provided using the Monte Carlo N-Particle Version 5 (MCNP5) code and continuous energy ENDF/B-V, VI.8, and VII.0, JEFF-3.1, and JENDL-3.3 nuclear data libraries.

Uncertainty Analyses with Nuclear Covariance Ddata in Reactor Core Ccalculations

Evaluated nuclear data are continuously improved. During the last years, the European library was updated from JEF-2.2 to JEFF-3.1 [1], the American library from ENDF/B-VI to ENDF/B VII [2], and the Japanese library from JENDL-3.2 to JENDL3.3/AC-2008 [3], with the main aim to achieve better agreement between calculated and measured results for integral systems. Along with this, growing attention is paid to uncertainty and sensitivity studies concerning the nuclear data evaluations, accompanied by improvements in the covariance data files describing the uncertainties of nuclear cross section data, and the calculation methods using these covariance data. For the validation of the nuclear data libraries, a large number of integral experiments are used. Descriptions of such experiments are found in the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” [4]. Most of these validation calculations refer to multiplication factors, although other measured quantities like reaction rates and reactivity coefficients are increasingly considered; such experiments are described in the “International Handbook of Evaluated Reactor Physics Benchmark Experiments” [5]. Most systems considered are compact assemblies, mainly at room temperature. Likewise, uncertainty and sensitivity investigations based on covariance data, as performed, e.g., with the TSUNAMI code package [6], primarily consider the multiplication factors of critical assemblies. Such compact critical systems at low temperatures are not necessarily representative for power reactors at operating conditions.

Consistent evaluation of modern nuclear data libraries for criticality safety analyses of thermal compound low-enriched-uranium systems

This paper addresses the assessment of MCNPX-2.5.0 criticality calculations using continuous-energy neutron cross-section libraries based on ENDF/B-VI.8, JEF-2.2, JENDL-3.3, and the recently released ENDF/B-VII.0 and JEFF-3.1 data files. For this purpose, validation calculations were performed for a suite of benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The benchmarks were selected on the basis of their similarity to designs found in today’s Light Water Reactor (LWR) compact storage pools and transport casks and in total comprised as much as 122 cases belonging to the category of thermal compound systems with low-enriched uranium. In order to define the ranges of applicability of the used calculational methods and to detect possible trends, the spectrum-related characteristics of the modeled critical experimental configurations were analysed. It is found that the latest ENDF/B-VII.0 and JEFF-3.1 libraries improve the older ENDF/B-VI.8 and JEF-2.2 compilations significantly and both lead to values for the weighted average of the normalized eigenvalues 〈 k eff calc / k eff exp 〉 very close to unity. All libraries have in common that any evidence of a statistically significant trend in the normalized eigenvalues versus both spectrum-related characteristics and experimental design parameters could not be found.

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.

KAMINI reactor benchmark analysis

Monte Carlo modeling of the Kalpakkam Mini reactor (KAMINI) had been made earlier by us for the first time using Monte Carlo code (MCNP4A) and ENDF/B-VI.2 data in the year 2004. KAMINI has been accepted as a part of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) recently and its detailed configuration is available for analysis in the International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2006. New neutron cross section data is also now available for nuclides of thorium–uranium fuel cycle through IAEA coordinated research and these are based on ENDF/B-VII. Examination of the new cross section data for nuclides 233U and 27Al, the major constituents in the fuel alloy of KAMINI, shows the addition of more resolved resonance structure in them. Hence, a review of our earlier work is made using new cross section data and by using the improved MCNP model of KAMINI supplied by ICSBEP evaluators. It is observed that by using the new data k eff value changed from 0.9890 to 0.99357 and is closer to the experimental value.

A Brief History of the Lawrence Radiation Laboratory Criticality Facility

Thousands of critical and high-multiplication subcritical experiments were performed at the Lawrence Radiation Laboratory (LRL) Critical Facility throughout the 1950s and 1960s. This paper presents a brief history of the origins of the LRL and the critical facility with a synopsis of those experiments that have been evaluated and published or will soon appear in the International Handbook of Evaluated Criticality Safety Benchmark Experiments.

Programs of Experiments with Critical Assemblies at the Russian Research Centre “Kurchatov Institute”

The paper gives a brief overview of benchmark experiments that have been performed and are being performed at the Russian Research Centre “Kurchatov Institutes” (RRC KIs), satisfy requirements of the International Criticality Safety Benchmark Evaluation Project (ICSBEP), and have been published or will be published in the “International Handbook of Evaluated Criticality Safety Benchmark Experiments.”These experiments include critical experiments in water-moderated facilities pertaining to substantiation of reactor physics for VVER-type light water reactors with uranium enrichments varying from natural uranium to ˜6.5%; in heterogeneous critical assemblies with a widely varying uranium enrichment (from 5 to 96%) for small nuclear power systems of various applications; in critical assemblies with a uranyl sulfate solution core; and in critical assemblies simulating peculiarities of high-temperature gas-cooled reactors (HTGR), RBMK physics, etc.A list of critical assemblies currently in operation at RRC KI is given. Future experimental programs are briefly described; their implementation, if based on the ICSBEP requirements, will be useful for the international community.Using RRC KI as an example, it is demonstrated that Russian nuclear centers maintain capabilities for carrying out a wide range of new critical experiments, including international cooperation in this area.

Use of International Criticality Safety Benchmark Evaluation Project Data for Validation of the Nuclear Data Library BAS

Effective neutron multiplication factors for 66 critical systems were calculated in order to test the neutron data library BAS. The class of systems chosen for the keff calculations includes unreflected metal uranium and plutonium systems and systems that were reflected by 238U, Fe, Al, Ti, Pb, Be, C, CH2, and H2O. Configurations and materials used in these critical systems were taken from the “International Handbook of Evaluated Criticality Safety Benchmark Experiments.” The calculations with BAS were performed using the codes PRIZMA-D and MCNP.4a. For comparison, the calculations were repeated using MCNP.4a with ENDF/B5 and ENDF/B6 cross-section data. A comparison of all results is provided.

Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the “International Handbook of Evaluated Criticality Safety Benchmark Experiments.” Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions.

Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the “International Handbook of Evaluated Criticality Safety Benchmark Experiments,” contributing to the preservation of a rather small number of heavy water benchmark critical experiments.

Use of International Criticality Safety Benchmark Evaluation Project Data for Validation of the ABBN Cross-Section Library with the MMK-KENO Code

The ABBN-93 299-group cross-section library is used as a source of neutron data for criticality calculations with the MMK-KENO Monte Carlo criticality code. Validation of criticality calculations is performed using the data presented in the “International Handbook of Evaluated Criticality Safety Benchmark Experiments.” This paper contains a description of a validation method based on statistical analysis of discrepancies between calculated and benchmark-model keff’s and the results of this validation for different types of experiments. Another validation method using a well-known procedure of group cross-section adjustment based on the maximum likelihood method (generalized least-squares method) and results of the validation for water-moderated highly enriched uranium homogeneous critical systems using selected experiments of the HEU-SOL-THERM type are also presented.

DICE: Database for the International Criticality Safety Benchmark Evaluation Program Handbook

The 2002 edition of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” (ICSBEP Handbook) spans more than 26 000 pages and contains 330 evaluations with benchmark specifications for 2881 critical or near-critical configurations. With such a large content, it became evident that the users needed more than a broad and qualitative classification of experiments to make efficient use of the ICSBEP Handbook. This paper describes the features of Database for the International Handbook of Evaluated Criticality Safety Benchmark Experiments (DICE), which is a database for the ICSBEP Handbook. The DICE program contains a relational database loaded with selected information from each configuration and a users’ interface that enables one to query the database and to extract specific parameters. Summary descriptions of each experimental configuration can also be obtained. In addition, plotting capabilities provide the means of comparing neutron spectra and sensitivity coefficients for a set of configurations.

The International Criticality Safety Benchmark Evaluation Project

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the U.S. Department of Energy. The ICSBEP became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Spain, and Israel are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled “International Handbook of Evaluated Criticality Safety Benchmark Experiments” (ICSBEP Handbook). The 2002 edition of the ICSBEP Handbook contains benchmark model specifications for 2881 critical or subcritical configurations that are intended for validating computer codes that calculate effective neutron multiplication and for testing basic nuclear data.

Use of International Criticality Safety Benchmark Evaluation Project Benchmarks for the Validation of CSAS26 (Scale 4.4a) for Configurations of Low-Enriched Uranium

As an example of the application of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) work to the nuclear industry, the validation of the control module CSAS26 of SCALE 4.4a for criticality calculations on a personal computer platform is presented. This work has been done using the models of critical experiments being compiled by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) since 1992. The description and results of this compilation were first presented during the Fifth International Conference on Nuclear Criticality Safety (ICNC’95). Out of 2881 critical configurations included in the latest edition (September 2002) of the ICSBEP “International Handbook of Evaluated Criticality Safety Benchmark Experiments,” NEA/NSC/DOC(95)03, OECD/NEA, 131 have been selected for the CSAS26 validation. The selected critical experiments have characteristics similar to the systems to be simulated with CSAS26 for low-enriched uranium (LEU) fuel fabrication applications. They represent both homogeneous configurations and hexagonally pitched rod lattices of low-enriched (from 1.60 to 5.00 wt% 235U) UO2 with several absorbers. The statistical uncertainties related to the application of CSAS26 for criticality calculations are also evaluated. The great number of cases involved allows an exhaustive statistical treatment of the data, including the analysis of correlations related to the type of system being simulated. The statistical uncertainties found are very small. As a result, the module CSAS26 is considered as a quite suitable calculational method for application to criticality safety analysis at LEU facilities.

The Benchmark Evaluation Process: From Experimental Data to Benchmark Model

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) provides a handbook of descriptions, evaluations, and models of experiments with fissionable material. The “International Handbook of Evaluated Criticality Safety Benchmark Experiments” (ICSBEP Handbook) is useful for criticality safety analysts and nuclear-data evaluators for validation of neutron transport codes and nuclear cross-section sets. Each of the four main parts of the ICSBEP document provides valuable information. The four parts are as follows: Part 1, detailed description of the experiment; Part 2, evaluation of experimental data to obtain parameter values that define the model and their uncertainties; Part 3, derivation and concise description of the benchmark model; and Part 4, sample calculation results. The ICSBEP Handbook provides a practical, standardized format for documenting nuclear experiments.Valuable, previously unknown data are often discovered during the evaluation process. Besides these discoveries, many other things have been learned during this first decade of evaluating and providing benchmark models of experiments. The current method is described in order to improve understanding of what is required to evaluate benchmark experiments for validation purposes.

A Historical Review of Critical Experiment Research Facilities in the United Kingdom

Experimental research facilities involving critical assemblies used in support of the U.K. nuclear power program are reviewed from the perspective of nuclear criticality safety. The scope of the U.K. program is outlined, and major facilities and programs are briefly reviewed. The methods and data arising from these programs, and their application for nuclear criticality safety assessment, are noted. The experiments selected for inclusion in the International Criticality Safety Benchmark Evaluation Project “International Handbook of Evaluated Criticality Safety Benchmark Experiments” are briefly outlined.