Reactor Physics Experiments Research Articles

Contributions to integral nuclear data in ICSBEP and IRPhEP since ND2013

The status of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) was last discussed directly with the international nuclear data community at ND2013. Since ND2013, integral benchmark data that are available for nuclear data testing has continued to increase. The status of the international benchmark efforts and the latest contributions to integral nuclear data for testing is discussed. Measurements found in the IRPhEP Handbook offer additional means to support integral nuclear data validation. Select benchmark configurations that have been added to the ICSBEP and IRPhEP Handbooks since ND2013 are highlighted.

Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a CockcrofteWalton type accelerator, which generates the external neutron source by deuteriumetritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

Neutron Characteristics of Solid Targets in Accelerator-Driven System with 100-MeV Protons at Kyoto University Critical Assembly

At the Kyoto University Critical Assembly, a series of reactor physics experiments on the accelerator-driven system (ADS) coupled with the fixed-field alternating gradient (FFAG) accelerator are carried out, and the spallation neutrons generated by 100-MeV protons from the FFAG accelerator are successfully injected into the cores. In the ADS experiments, the neutron characteristics of the solid target are investigated through static and kinetic analyses, when the external neutron source of the neutron spectrum (the W, W-Be, or Pb-Bi target) is varied. The results demonstrate that the neutron yield is large with the W target, but a discrepancy is observed between the experiments and the calculations, because the experimental uncertainty of proton monitoring is attributable to defocusing of proton beams. With the use of reaction rate distribution in the core region, the static parameters are estimated fairly well in the analyses of the neutron multiplication and subcritical multiplication factor. In the kinetic experiments, the variation of the solid target used is clearly evident in the prompt neutron decay constant and the subcriticality.

Preface: Special Issue on the International Reactor Physics Experiment Evaluation Project

Preface: Special Issue on the International Reactor Physics Experiment Evaluation Project

Reactivity Worth Measurement of Major Fission Products in MINERVE LWR Lattice Experiment

The CERES collaborative program between the French Atomic Energy Commission (CEA) and the United Kingdom Atomic Energy Agency (UKAEA) was performed in the MINERVE and DIMPLE reactors at Cadarache and Winfrith, respectively. CERES Phase II was devoted to the validation of fission product (FP) poisoning through the reactivity worth measurements of FP samples. This paper describes the oscillation experiment at the center of the MINERVE pressurized water reactor–type test lattice. This experiment is strongly representative of the FP poisoning in light water reactor spent fuels because the separated FP isotope is introduced into real UO2 pellets where 238U/iFP resonance overlap occurs. In order to preserve the experimental results within the International Reactor Physics Experiment Evaluation international database, slight corrections to define a two-dimensional benchmark are presented. The evaluation of experimental uncertainties is detailed. Therefore, the accurate APOLLO2.8 analysis of this benchmark is described, using recent JEFF-3.1.1 and ENDF/B-VII.0 nuclear data files. JEFF-3.1.1 FP worth is particularly satisfactory, except for 153Eu, which is underestimated by 8.1% ± 2.6%. The CERES integral measurement data also suggest improvements to 99Tc and 145Nd evaluations in the ENDF/B-VII.0 library.

IRPhEP Evaluation of VENUS 9 and VENUS 7 Experiments

The zero-power reactor VENUS (Vulcain Experimental Nuclear Study) was built in 1963–1964 at CEN-SCK, Mol, Belgium, as a nuclear mock-up of a projected spectral shift marine reactor called VULCAIN. The facility was modified in 1966 and 1967 in preparation for carrying out a series of critical experiments for the Belgian Plutonium Recycle Programme (PRP), which was partially supported by EURATOM. This was the VENUS-PRP program that took place between 1967 and 1975. VENUS-PRP-9 and VENUS-PRP-9/1, and VENUS-PRP-7 were two series of these PRP configurations that were carried out in 1967–1968 and that have recently been subject to evaluations as part of the Organisation for Economic Co-operation and Development, Nuclear Energy Agency International Reactor Physics Experiment Evaluation Project (IRPhEP). The VENUS-PRP-9 and VENUS-PRP-9/1 configurations focused on the study of the power distribution across the boundary between a standard UO2 fuel region, enriched to 4 wt% 235U, and a mixed oxide fuel region made of UO2, enriched to 3 wt% 235U with ˜1 wt% PuO2, simulating a one-cycle burnt fuel. The IRPhEP evaluation focused on evaluating reaction rates and powers measured along a line that crossed the boundary between the two regions. In the VENUS-PRP-7, VENUS-PRP-7/1, and VENUS-PRP-7/3 series—which used essentially the same fuel pins—reactivities, reactivity worths of substituted and removed fuel pins, and radial fission rate distributions were measured; these quantities were evaluated in the framework of the IRPhEP project.

Evaluation of the KRITZ-2 Criticality and Reaction Rate Benchmark Experiments

A complete evaluation of the experimental uncertainties of the KRITZ-2 series of critical and relative fission rate experiments was performed within the International Reactor Physics Experiment Evaluation Project. The uncertainties in the benchmark model keff are mainly due to uranium enrichment, plutonium content [mixed oxide (MOX) fuel], pitch, and boron isotopic composition. The largest contribution to the uncertainty in the benchmark model keff is from the uncertainty in the bias due to the homogenization of the particulate MOX fuel. In addition, uncertainties due to nuclear data libraries are presented.The keff’s calculated with various nuclear data libraries systematically underpredict the benchmark model keff by one to three times the standard experimental uncertainties. When taking into account uncertainties in nuclear data estimated using SCALE-6.0 and JENDL-4.0m covariances, the benchmark and calculated keff’s agree within 1σ of the total—experimental plus calculational—uncertainties. In contrast to the criticality benchmark data, the calculated relative fission rates agree very well with the experimental ones, especially when eliminating systematic errors due to normalization.

The Evaluation of the Effective Kinetic Parameters and Reactivity of the IPEN/MB-01 Reactor for the International Reactor Physics Experiment Evaluation Project

The evaluation of the experiments of the effective delayed neutron parameters and reactivity performed in the IPEN/MB-01 research reactor facility has been successfully accomplished. The evaluated data are of very good quality and fulfill the requirements of a benchmark. The recently released MCNP6 together with the ENDF/B-VII.1, JENDL-4.0, JEFF-3.1.1, ENDF/B-VII.0, and JENDL-3.3 nuclear data libraries has been employed to calculate the effective delayed neutron parameters adopting the benchmark model of the IPEN/MB-01 reactor available in the International Handbook of Reactor Physics Benchmark Experiments. The analysis reveals that all these nuclear data libraries produced satisfactory results for ²eff, ²eff/ᴧ, and ᴧ. The same cannot be said for determining the reactivity using the Inhour equation. It was shown that there is a clear tendency to increase the deviation with the absolute value of the reactivity for negative periods. Only JENDL-3.3 and JEFF-3.1.1 produced results that are inside the 3σ range of the benchmark value uncertainty. Specifically for the case of ENDF/B-VII.1, a good part of this discrepancy is due to the decay constant of the first group of delayed neutrons, which is overestimated according to the experimental value measured in the IPEN/MB-01 reactor.

Applications of Integral Benchmark Data

The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) provide evaluated integral benchmark data that may be used for validation of reactor physics/nuclear criticality safety analytical methods and data, nuclear data testing, advanced modeling and simulation, and safety analysis licensing activities. The handbooks produced by these programs are used in over 30 countries. Five example applications are presented in this paper: (a) use of IRPhEP data in uncertainty analyses and cross-section adjustment, (b) uncertainty evaluation methods for reactor core design at Japan Atomic Energy Agency using reactor physics experimental data, (c) application of benchmarking data to a broad range of criticality safety problems, (d) cross-section data testing with ICSBEP benchmarks, and (e) use of the International Handbook of Evaluated Reactor Physics Benchmark Experiments to support the power industry.

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.

Uncertainty Analysis of Temperature Effect on Reactivity in Very High Temperature Reactor Critical Assembly and Establishment of Biases for Benchmark Models

A series of integral reactor physics experiments conducted at the Very High Temperature Reactor Critical Assembly was analyzed to assemble it into a benchmark through an extensive peer-review process under the International Reactor Physics Experiment Evaluation Project. This benchmark provides the experimental data with respect to the criticalities of seven core configurations and the temperature effect on reactivity up to 200°C with explicit experimental uncertainties newly evaluated. It further presents the benchmark models and corresponding values with some simplifications so that it can be used by reactor designers for validating the analytical tools employed to design next-generation reactors and for establishing the safety basis for operation of these reactors.

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.

Evaluation of Burnup Reactivity Coefficients Measured in Experimental Fast Reactor JOYO MK-I Duty Power Operation Cycles

To provide a reactor physics benchmark for burnup reactivity coefficients, experimental data, showing the relationship between excess reactivity and accumulated thermal power acquired during the experimental fast reactor JOYO MK-I duty power operation in the late 1970s, have been evaluated and analyzed. To improve the prediction accuracy of nuclear characteristics through the use of integral experimental data, nominal values and uncertainties, including correlations of the experimental data, were evaluated. All possible uncertainty factors were evaluated and quantified by utilizing knowledge obtained after the MK-I duty power operation and calculation results based on the latest reactor physics analysis methods. Meanwhile, the present evaluated data have been reviewed and approved by the International Reactor Physics Experiment Evaluation Project, with the expectation that these data will be widely used. In the present paper, the evaluation of nominal values and uncertainties is described with a focus on the measurement technique uncertainty, which is a dominant uncertainty factor of the burnup reactivity coefficient. In addition, new analysis results of the benchmark problem are shown by the use of the latest Japanese evaluated nuclear data JENDL-4.0.

Preface: Special Issue on the International Reactor Physics Experiment Evaluation Project

Preface: Special Issue on the International Reactor Physics Experiment Evaluation Project

ZPPR Benchmarks for Large LMFBR Core Physics from JUPITER Cooperative Program Between United States and Japan

Under the International Reactor Physics Experiment Evaluation Project (IRPhEP) framework, in the cooperative JUPITER program between the United States and Japan, benchmarks are established to study large fast breeder reactor (FBR) core physics utilizing nine Zero Power Plutonium Reactor (ZPPR) critical experimental cores. These benchmarks cover a wide variety of core concepts including homogeneous and heterogeneous configurations, clean and engineering mock-up cores of 600- to 1000-MW(electric)–class sizes, and various core parameters such as criticality, reaction rate, and reactivity. Recently, detailed experimental information from original documents from Argonne National Laboratory has been scrutinized very carefully to establish the benchmark model and to evaluate quantitatively the experimental uncertainty. The benchmarks supply users with heterogeneous cell models and three-dimensional (3-D) core configurations, which are simplified to a degree that preserves the important physical features of the ZPPR cores such as plate heterogeneity, different drawer types, and 3-D core arrangement. Further, the benchmark handbook includes as-built information of the ZPPR cores as a complete set of electronic form; therefore, a user can develop his or her own benchmark model if necessary. The analysis of the benchmark with the deterministic or Monte Carlo method demonstrates its usefulness both for improving analytical methods and for validating nuclear data.

An Overview of the International Reactor Physics Experiment Evaluation Project

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next-generation reactor and advanced fuel cycle concepts. Two Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

MANTRA: An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

This paper presents an update of an on-going collaborative INL-ANL-ISU integral reactor physics experiment whose objective is to infer the effective neutron capture cross sections for most of the actinides of importance for reactor physics and fuel cycle studies in both fast and epithermal spectra. Some fission products are also being considered. The principle of the experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation together with the neutron fluence will allow inference of effective neutron capture cross-sections in different neutron spectra.

Integral Benchmark Data for Nuclear Data Testing Through the ICSBEP & IRPhEP

The status of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and International Reactor Physics Experiment Evaluation Project (IRPhEP) was last discussed directly with the nuclear data community at ND2007. Since ND2007, integral benchmark data that are available for nuclear data testing have increased significantly. The status of the ICSBEP and the IRPhEP is discussed and selected benchmark configurations that have been added to the ICSBEP and IRPhEP Handbooks since ND2007 are highlighted.

Monte Carlo and deterministic computational methods for the calculation of the effective delayed neutron fraction

The effective delayed neutron fraction βeff plays an important role in kinetics and static analysis of the reactor physics experiments. It is used as reactivity unit referred to as “dollar”. Usually, it is obtained by computer simulation due to the difficulty in measuring it experimentally. In 1965, Keepin proposed a method, widely used in the literature, for the calculation of the effective delayed neutron fraction βeff. This method requires calculation of the adjoint neutron flux as a weighting function of the phase space inner products and is easy to implement by deterministic codes. With Monte Carlo codes, the solution of the adjoint neutron transport equation is much more difficult because of the continuous-energy treatment of nuclear data. Consequently, alternative methods, which do not require the explicit calculation of the adjoint neutron flux, have been proposed. In 1997, Bretscher introduced the k-ratio method for calculating the effective delayed neutron fraction; this method is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Using Monte Carlo calculation Bretscher evaluated the βeff as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as the k-ratio method). In the present work, the k-ratio method is applied by Monte Carlo (MCNPX) and deterministic (PARTISN) codes. In the latter case, the ENDF/B nuclear data library of the fuel isotopes (235U and 238U) has been processed by the NJOY code with and without the delayed neutron data to prepare multi-group WIMSD neutron libraries for the lattice physics code DRAGON, which was used to generate the PARTISN macroscopic cross sections. In recent years Meulekamp and van der Marck in 2006 and Nauchi and Kameyama in 2005 proposed new methods for the effective delayed neutron fraction calculation with only one Monte Carlo computer simulation, compared with the k-ratio method which require two criticality calculations. In this paper, the Meulekamp/Marck and Nauchi/Kameyama methods are applied for the first time by the MCNPX computer code and the results obtained by all different methods are compared.

Evaluation and Uncertainty Analysis of the KRITZ-2 Critical Benchmark Experiments

To validate new nuclear cross-section evaluations and computational methods, a large number of benchmark experiments were performed in the past. The Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) launched several projects aiming to collect, preserve, and disseminate the benchmark data in a user-friendly format. Reactor physics benchmarks are covered by the International Reactor Physics Experiments (IRPhE) project. This paper presents the preparation of the IRPhE compilation for the KRITZ-2 critical experiments, consisting of altogether six configurations, both UO2 and mixed oxide, measured at two different temperatures at Studsvik. These configurations were selected for the purpose of the OECD/NEA uncertainty analysis in modeling benchmark activities. Uncertainties due to input data uncertainties, modeling errors, and numerical approximations were studied, with particular emphasis on the uncertainties in the nuclear cross-section data. The SUSD3D sensitivity-uncertainty code with the SCALE-6.0, JENDL-4, and/or JENDL-3.2 covariance data were used in this study. The consistency among the calculated-to-experiment values and the overall computational uncertainties is discussed.