Fission Rate Measurements Research Articles

Measurement of the α ratio and (n,γ) cross section of U235 from 0.2 to 200 eV at n_TOF

We measured the neutron capture-to-fission cross-section ratio (α ratio) and the capture cross section of $^{235}$U between 0.2 and 200 eV at the n_TOF facility at CERN. The simultaneous measurement of neutron-induced capture and fission rates was performed by means of the n_TOF BaF$_{2}$ Total Absorption Calorimeter (TAC), used for detection of γ rays, in combination with a set of micromegas detectors used as fission tagging detectors. The energy dependence of the capture cross section was obtained with help of the $^{6}$Li(n,t) standard reaction determining the n_TOF neutron fluence; the well-known integral of the $^{235}$U(n,f) cross section between 7.8 and 11 eV was then used for its absolute normalization. The α ratio, obtained with slightly higher statistical fluctuations, was determined directly, without need for any reference cross section. To perform the analysis of this measurement we developed a new methodology to correct the experimentally observed effect that the probabilities of detecting a fission reaction in the TAC and the micromegas detectors are not independent. The results of this work have been used in a new evaluation of $^{235}$U performed within the scope of the Collaborative International Evaluated Library Organisation (CIELO) Project, and are consistent with the ENDF/B-VIII.0 and JEFF-3.3 capture cross sections below 4 eV and above 100 eV. However, the measured capture cross section is on average 10% larger between 4 and 100 eV.

Validation of Pin-Resolved Reaction Rates, Kinetics Parameters, and Linear Source MOC in MPACT

Over the years, significant validation work for the neutronics code MPACT has been performed against zero-power critical benchmarks and measured data from operating nuclear power plants. Among all of these efforts, however, validation of the pin-resolved capability in MPACT has been limited by the public availability of experimental data and to a lesser degree availability of measurement techniques and facilities that provide such detailed data. Recently, new measurement results to experimentally determine the reaction rate along the pellet radius from the IPEN/MB-01 research reactor facility (IPEN) have been published as a benchmark in the International Reactor Physics Experiment Project handbook. In this paper, we examine MPACT simulation results for several IPEN benchmark experiments with emphasis on the intrapin reaction rate measurements. The IPEN critical experiments with variations in system temperature and gadolinium loadings are modeled first with the latest MPACT cross-section library and linear source (LS) method of characteristics (MOC) capability. The MPACT results of two-dimensional (2-D) models with axial buckling are within 160 pcm from the experimental eigenvalues using the flat source MOC. Using the LS MOC, the errors are no more than 70 pcm, and the temperature trend of various cases is smaller. The MPACT three-dimensional models with LS show slightly worse comparisons than the 2-D models, which may be due to the isotropic transverse leakage and homogenized cross-section approximations of the 2-D/one-dimensional solver. For the reaction rate validation, MPACT produces intrapin reaction rate results within 2σ of the experiment and shows excellent agreement with the Monte Carlo solution. The observed discrepancies between the simulated results and experiment for the fission rate measurements are discussed. The kinetics parameters measured in another IPEN experiment are also compared with MPACT simulations using different kinetics data sources. According to the validation results, JENDL-4.0 and Santamarina et al.’s data are recommended for MPACT transient calculations.

Computational validation of the fission rate distribution experimental benchmark at the JSI TRIGA Mark II research reactor using the Monte Carlo method

A fission rate profile benchmark experiment has been performed at the Jožef Stefan Institute TRIGA Mark II reactor. The measurements were made using absolutely calibrated miniature fission chambers developed and manufactured by the Commissariat à l’Énergie Atomique et aux Énergies Alternatives. The aim of the paper is to describe the experimental set-up, fission rate measurements and to present the detailed Monte Carlo computational model of the TRIGA reactor, which was constructed with as used to compute absolute fission rate distributions in the core at a fixed control rod position, taking into account the detailed description of the experimental configuration. The paper focuses on the extensive evaluation of experimental and calculational uncertainties and biases following the International Reactor Physics Experiment Evaluation Project methodology. A comparison between the measured and computed absolute reaction rates concludes the paper, with the agreement being within one sigma standard uncertainty.

Uncertainty quantification and representativity analysis of LWR-PROTEUS Phase III experiments using SHARKX

SHARKX is a set of Perl-based tools build around the lattice code CASMO-5, used to perform sensitivity analysis (SA), uncertainty quantification (UQ) and representativity analysis (RA) developed at the Paul Scherrer Institut. The paper describes the application of recent developments in the SHARKX methodology to zero-power reactor experiments, namely the LWR-PROTEUS Phase III (LWR-III) experiments.The SHARKX Stochastic Sampling (SS) pinpower UQ method is applied to the LWR-III campaign, which featured SVEA-96 Optima2 fresh BWR fuel assemblies, in an attempt to explain the small discrepancy previously observed between the fission rate measurements (E) and their predictions with CASMO-5 (C). The pinpower uncertainty due to nuclear data and technological parameters is in the range of 0.4–0.8%. This additional source of uncertainty, combined with the 0.6% measurement uncertainty, failed to fully explain the remaining few C/E discrepancies in the pinpower map. The source of nuclear data and technological parameters uncertainty is either underestimated, not well modeled or an additional source of uncertainty has not been taken into account.The relevance of the LWR-III experiments to the usage of SVEA-96 Optima2 assemblies in a BWR is rigorously assessed using the newly developed RA capability of SHARKX. The pinpower maps of the three different radial cross-sections of the SVEA-96 Optima2, featuring 96, 92 and 84 pins, are measured in turn in the LWR-III experiments; and various thermal–hydraulic conditions encountered during normal operation of a BWR are considered for the applications. The representativity coefficients ck between each radial map of the LWR-III experiments and the considered applications are always higher than 0.9. It shows that the LWR-III experiments are highly representative of the applications considered. The representativity decreases with increasing void fraction in the application configuration, which is expected as the amount of D2O in the PROTEUS test tank during the LWR-III experiment is designed to simulate moderation of a HFP BWR at 0% void. The influence of the Variance Covariance Matrix origin on the representativity results is responsible for up to 3.5% variation in ck coefficient for experiments and applications with lower ck values.

Measurement and analysis of fission rates in depleted uranium and depleted uranium–polyethylene shells with neutrons from D-T reaction

To validate the concept design of a novel fusion–fission hybrid energy reactor, a depleted uranium assembly and a combined assembly of uranium and polyethylene were designed and assembled based on a depleted uranium spherical shell and a polyethylene spherical shell. The distribution of the fission rates for the depleted uranium and enriched uranium in the two assemblies, as a function of the distance of the detection position to the centre, was measured using a plate fission chamber bombarded by D-T neutrons. The addition of a polyethylene shell significantly changed the neutron spectrum; in particular, the neutron fluxes with energies of 1 MeV and lower were changed. Using MCNP5 and the attached libraries, the fission rate experiments were simulated, and the experimental configuration, including the wall of the experimental hall, was described in detail in the model. The fission rate distributions for depleted uranium and enriched uranium in the two assemblies were reproducible. The difference between the calculated results with different libraries and different tallies is as small as 1.0%. By considering the neutron flux, the fission rate and the C/E values, it is concluded that the fission rates of depleted uranium and enriched uranium induced by the fast neutrons were overestimated, and it is proposed that the fission parameters of uranium for fast neutrons should be re-evaluated, or the margin of the concept design should be enlarged, to make the concept effective.

Measurement and analysis of fission rates in a spherical mockup of uranium and polyethylene

Measurements of the reaction rate distribution were carried out using two kinds of Plate Micro Fission Chamber (PMFC). The first is a depleted uranium chamber and the second an enriched uranium chamber. The material in the depleted uranium chamber is strictly the same as the material in the uranium assembly. With the equation solution to conduct the isotope contribution correction, the fission rate of 238U and 235U were obtained from the fission rate of depleted uranium and enriched uranium. Then, the fission count of 238U and 235U in an individual uranium shell was obtained. In this work, MCNP5 and continuous energy cross sections ENDF/BV.0 were used for the analysis of fission rate distribution and fission count. The calculated results were compared with the experimental ones. The calculation of fission rate of DU and EU were found to agree with the measured ones within 10% except at the positions in polyethylene region and the two positions near the outer surface. Because the fission chamber was not considered in the calculation of the fission counts of 238U and 235U, the calculated results did not agree well with the experimental ones.

Experimental Validation of 3-D Reconstructed Pin Power Distributions in Full-Scale BWR Fuel Assemblies with Partial-Length Rods

Total fission rate measurements have been performed on full-size boiling water reactor fuel assemblies of type SVEA-96 Optima2 in the framework of phase III of the light water reactor (LWR)-PROTEUS experimental program at Paul Scherrer Institute. This paper presents comparisons of calculated, nodal reconstructed, pinwise total fission rate distributions with experimental results. Radial comparisons have been performed for the three axial sections of the assembly (96, 92, and 84 fuel pins), while three-dimensional (3-D) effects have been investigated at pellet level for the two transition regions, i.e., the tips of the short (one-third) and long (two-thirds) partial-length rods. The test zone has been modeled using two different code systems: HELIOS/PRESTO-2 and CASMO-5/SIMULATE-5. The former is presently used for core monitoring and design at the Leibstadt Nuclear Power Plant (KKL). The latter represents the most recent generation of codes constituting the widely applied CASMO/SIMULATE system. For representing the PROTEUS test zone boundaries, partial current ratios—derived from a 3-D Monte Carlo (MCNPX) model of the entire reactor—have been applied to the PRESTO-2 and SIMULATE-5 models in the form of two-group and five-group diagonal albedo matrices, respectively. The MCNPX results have also served as a reference high-order transport solution in the calculation-to-experiment (C/E) comparisons.It is shown that the performance of the nodal methodologies in predicting the global distribution of the total fission rate is very satisfactory. Considering the various radial comparisons, the standard deviations of the C/E distributions do not exceed 1.9% for any of the three methodologies—PRESTO-2, SIMULATE-5, and MCNPX. For the 3-D comparisons at pellet level, the corresponding standard deviations are 2.7%, 2.0%, and 2.1%, respectively.

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [M. B. Chadwick, P. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Br, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, and S. C. van der Marck, “ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,” Nuclear Data Sheets 107, 2931 (2006)].

CREOLE experiment study on the reactivity temperature coefficient with sensitivity and uncertainty analysis using the MCNP5 code and different neutron cross section evaluations

In the present work, we analyze the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA/Cadarache, was mainly dedicated to the RTC studies for both UO 2 and UO 2–PuO 2 PWR type lattices covering the whole temperature range from 20 °C to 300 °C. We have developed an accurate 3D model of the EOLE reactor by using the MCNP5 Monte Carlo code which guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII, JEFF3.1 and JENDL3.3 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values; the discrepancy is less than 0.23 pcm/°C. Whereas, when using ENDF-BVII evaluation for UOX configuration, this discrepancy reaches a value of 0.63 pcm/°C. In order to specify the source of this relatively large discrepancy in the case of ENDF-BVII nuclear data evaluation, the k eff discrepancy between ENDF-BVII and JENDL3.3 was decomposed by using sensitivity and uncertainty analysis technique. The thermal energy range of 238U( n,γ) absorption cross section was found to contribute to the major part of the observed k eff discrepancy between ENDFB-VII and JENDL3.3 evaluations.

Experimental validation of radial reconstructed pin-power distributions in full-scale BWR fuel assemblies with and without control blade

Total fission rate measurements have been performed on full-size BWR fuel assemblies of type SVEA-96+ in the zero power reactor PROTEUS at the Paul Scherrer Institute. This paper presents comparisons of reconstructed 2D pin fission rates from nodal diffusion calculations to the experimental results in two configurations: one “regular” (I-1A) and the other “controlled” (I-2A). Both configurations consist of an array of 3 × 3 SVEA-96+ fuel assemblies moderated with light water at 20 °C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the north-west corner of the central fuel assembly. To minimise the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3 × 3 experimental configuration (test zone) was modelled using the core monitoring and design tools that are applied at the Leibstadt Nuclear Power Plant (KKL). These are the 2D transport code HELIOS, used for the cross-section generation, and the 3D, 2-group nodal diffusion code PRESTO-2. The exterior is represented, in the axial and radial directions, by 2-group partial current ratios (PCRs) calculated at the test zone boundary using a 3D Monte Carlo (MCNPX) model of the whole PROTEUS reactor. Sensitivity cases are analysed to show the impact of changes in the 2D lattice modelling on the calculated fission rate distribution and reactivity. Further, the effects of variations in the test zone boundary PCRs and their behaviour in energy are investigated. For the test zone configuration without control blade, the pin-power reconstruction methodology delivers the same level of accuracy as the 2D transport calculations. On the other hand, larger deviations that are inherent to the use of reflected geometry in the lattice calculations are observed for the configuration with the control blade inserted. In the basic (reference) simulation cases, the calculated-to-experimental ( C/ E) ratios of the total fission rate have a standard deviation of 1.3% in configuration I-1A (regular) and 3.2% in configuration I-2A (controlled). These results are only slightly affected by changes in the lattice modelling. On the other hand, the use of correction factors to account for the spatial variation of the test zone PCRs tends to improve the C/ E ratios, mainly in the controlled case.

Measurements of the number of neutrons emitted per fission in a fast neutron spectrum for 235U, 237Np, and 243Am

Experiments were designed and performed to determine the number of neutrons emitted per fission, v, for 235U, 237Np and 243Am in a fast neutron spectrum. The neutron emission rates were observed using three proton recoil detectors. The detection system's efficiency was determined using a 252Cf source. The fission rate measurements were performed using solid state nuclear track detectors (SSNTD). The measured values of v in the fast spectrum of the 144 keV silicon-filtered neutron beam for 235U, 237Np, and 243Am are 2.54 ± 0.06, 3.13 ± 0.12, and 4.00 ± 0.35, respectively.

Experimental Study of the Large-Scale Axially Heterogeneous Liquid-Metal Fast Breeder Reactor at the Fast Critical Assembly —Power Distribution Measurements and Their Analyses

Power distributions of the large-scale axially heterogeneous liquid-metal fast breeder reactor were studied by using the experiment results of fast critical assemblies XI, XII, and XIII and the results of their analyses. The power distributions were examined by the gamma-scanning method and fission rate measurements using /sup 239/Pu and /sup 238/U fission counters and the foil irradiation method. In addition to the measurements in the reference core, the power distributions were measured in the core with a control rod inserted and in a modified core where the shape of the internal blanket was determined by the radial boundary. The calculation was made by using JENDL-2 and the Japan Atomic Energy Research Institute's standard calculation system for fast reactor neutronics. The power flattening trend, caused by the decrease of the fast neutron flux, was observed in the axial and radial power distributions. The effect of the radial boundary shape of the internal blanket on the power distribution was determined in the core. The thickness of the internal blanket was reduced at its radial boundary. The influence of the internal blanket was observed in the power distributions in the core with a control rod inserted. The calculation predicted the neutron spectrum hardermore » in the internal blanket. In the radial distributions of /sup 239/Pu fission rates, the space dependency of the calculated-to-experiment values was found at the active core close to the internal blanket.« less

Fission Rates and High-Energy Neutron Fluxes in the Purdue Fast Breeder Blanket Facility

Fission rate measurements were performed in the Purdue Fast Breeder Blanket Facility (FBBF) and compared with two-dimensional 50-group diffusion calculations on an absolute basis. Fission rates in /sup 239/Pu, /sup 235/U, /sup 237/Np, /sup 238/U, and /sup 232/Th were measured using fused quartz fission track recorders. Calculations using the 1DX and 2DB codes, and the LIB-IV nuclear data library were compared with the measurements in the form of reaction rate calculated-to-experiment (C/E) trajectories. The investigations were aimed at providing an improved understanding and description of the high-energy neutron fluxes and reaction rate distributions. Also investigated were previously reported C/E reaction rate discrepancies in the blanket.

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Solid state track recorder fission rate measurements in low power light water reactor pressure vessel mockups

Abstract Solid State Track Recorder (SSTR) measurements of neutron-induced fission rates have been made in several pressure vessel mockup facilities as part of the U.S. Nuclear Regulatory Commission sponsored Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program. These measurements have been made at U.S., Belgian, and British facilities which have been designed to replicate various aspects of light water reactor core, water gap, thermal shield, pressure vessel and reactor cavity geometries. The results of SSTR fission rate measurements, comparisons of these measurements with other techniques, and comparisons to theoretical calculations are presented.

Solid state track recorder fission rate measurements at high neutron fluence and high temperature

Solid State Track Recorder (SSTR) techniques have been used to measure 239 Pu, 235 U, and 237 Np fission rates for total neutron fluences approaching 5 × 10 17 n cm -2 at temperatures in the range 680–830°F. Natural quartz crystal SSTRs were used to withstand the high temperature environment, and ultra low-mass fissionable deposits of the three isotopes were required to yield scannable track densities at the high neutron fluences. The results of these high temperature, high neutron fluence measurements are reported.

Neutron Fission Rate Measurements across Poisoning Discontinuity in Water

By making use of miniature 239Pu and 235U fission chambers and an accelerator-based neutron source, the 239Pu/235U neutron fission rate ratio profile has been measured in the region of a plane interface between a tank containing ordinary water and a tank containing an aqueous solution of boric acid. The reported preliminary results indicate an approximately 8 mm shift of the central point of the fission rate ratio profile towards the poisoned side of the system and an accompanying larger shift in the 235U Cd ratio profile towards the unpoisoned side.

Neutron fission rate measurements across poisoning discontinuity in water.

By making use of miniature 239Pu and 235U fission chambers and an accelerator-based neutron source, the 239Pu/235U neutron fission rate ratio profile has been measured in the region of a plane interface between a tank containing ordinary water and a tank containing an aqueous solution of boric acid. The reported preliminary results indicate an approximately 8 mm shift of the central point of the fission rate ratio profile towards the poisoned side of the system and an accompanying larger shift in the 235U Cd ratio profile towards the unpoisoned side.

Fission Rate Measurements with the Makrofol Solid-State Track Recorder

Fission rates have been experimentally determined in the NISUS standard fast assembly, installed on the University of London Reactor CONSORT, with a Makrofol solid-state track recorder (SSTR). The ...

Prompt [formula omitted] of 235U and 239Pu in some fast reactor spectra

The averaged number of prompt neutrons per 335sU and 239Pu fission have been measured in fast and thermalized neutron fluxes at the fast zero-power reactor (FRO) at Studsvik, Sweden. The measurements were made with three different reactor cores giving median fission energies for 235U (core centre) of 50, 170 and 240 keV. The spectrum averaged v data were compared with the corresponding thermal ν data. The ν- value of the spontaneous fission of 252Cf was used as an intermediate standard. The fission samples were placed in neutron beams extracted from the reactor cores and a large liquid scintillator was used as the fission neutron detector. The experiment also included relative fission rate measurements, 238U/ 235U, for the three reactor cores. The results are compared with calculated relative ν- values deduced from different sets of recently evaluated fission cross section and ν- data and from different assumptions concerning the reactor neutron spectra. Within the experimental one standard deviation errors of 1·5 and 2 per cent for 235U and 239Pu a good agreement was obtained between the experimental and calculated data assuming ν- values in the fission resonance region equal to ν at thermal energy.