Kyoto University Critical Assembly Research Articles

Sensitivity of a control rod worth estimate to neutron detector position by time-dependent Monte Carlo simulations of the rod drop experiment

The control rod worth sensitivity to the neutron detector position in the rod drop experiment is studied by the time-dependent Monte Carlo (TDMC) neutron transport calculations for AGN-201K educational reactor and the Kyoto University Critical Assembly. The TDMC simulations of the rod drop experiments are conducted by the Seoul National University Monte Carlo (MC) code, McCARD, yielding time-dependent neutron densities at detector positions. The detector-position-dependent results of the total control rod worth calculated by the extrapolation, the integral counting, and the inverse methods are compared with the numerical reference using the MC eigenvalue calculations and the experimental results. From these comparisons, it is observed that the total control rod worth can be estimated with a considerable difference depending on the detector position through the rod drop experiment. The proposed TDMC simulation of the rod drop experiment can be applied for searching a better detector position or quantifying a bias for the control rod worth measurement.

Impact of Uncertainty Reduction on Lead-Bismuth Coolant in Accelerator-Driven System Using Sample Reactivity Experiments

In this study, we have demonstrated that data assimilation (DA) using lead and bismuth sample reactivities measured in the Kyoto University Critical Assembly A-core can successfully reduce the uncertainty of the coolant void reactivity in accelerator-driven systems (ADSs) derived from inelastic scattering cross sections of lead and bismuth. We reevaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the DA formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties and performed DA using the reactor analysis code system MARBLE. The high-sensitivity coefficients of the sample reactivities to lead and bismuth allowed us to reduce the cross-section–induced uncertainty of the void reactivity of the ADS from 6.3% to 4.8%, achieving a provisional target accuracy of 5% in this study. Furthermore, we demonstrated that the uncertainties arising from other dominant factors, such as minor actinides and steel, can be effectively reduced by using integral experimental data sets for the unified cross-section dataset ADJ2017.

Experimental Analyses of Capture Reaction Rates for Epithermal and Resonance Neutrons in Source-Driven Subcritical Cores

Neutron irradiation experiments are carried out in source-driven subcritical cores with high-energy neutrons generated by spallation reactions between a 100-MeV proton beam and a lead-bismuth target at the Kyoto University Critical Assembly. The main objective of the experiments is to investigate the effect of epithermal and resonance neutrons on the accuracy of capture reaction rates with respect to a subcriticality variation. Activation foils of copper, indium, tantalum, and tungsten are employed to obtain capture reaction rates for epithermal and resonance neutrons by applying the cadmium difference method. Also, the applicability of the foils for the measurement of the reaction rates for epithermal and resonance neutrons is substantiated in the critical irradiation experiments performed prior to the subcritical experiments. The subcritical experiments are conducted with three different subcriticalities by changing the control rod insertion pattern. The measured reaction rates are compared with the calculated values obtained by the Monte Carlo code MVP with JENDL-4.0, and the ratio of the calculation and experiment values of the reaction rates shows equivalent values within the 1σ errors regardless of a difference in the subcriticality. The compared results indicate that the numerical analyses have a consistent accuracy of reaction rates in epithermal and resonance energy regions for a subcriticality variation in source-driven subcritical cores.

Advances for the time-dependent Monte Carlo neutron transport analysis in McCARD

For an accurate and efficient time-dependent Monte Carlo (TDMC) neutron transport analysis, several advanced methods are newly developed and implemented in the Seoul National University Monte Carlo code, McCARD. For an efficient control of the neutron population, a dynamic weight window method is devised to adjust the weight bounds of the implicit capture in the time bin-by-bin TDMC simulations. A moving geometry module is developed to model a continuous insertion or withdrawal of a control rod. Especially, the history-based batch method for the TDMC calculations is developed to predict the unbiased variance of a bin-wise mean estimate. The developed methods are verified for three-dimensional problems in the C5G7-TD benchmark, showing good agreements with results from a deterministic neutron transport analysis code, nTRACER, within the statistical uncertainty bounds. In addition, the TDMC analysis capability implemented in McCARD is demonstrated to search the optimum detector positions for the pulsed-neutron-source experiments in the Kyoto University Critical Assembly and AGN-201K.

Void Reactivity in Lead and Bismuth Sample Reactivity Experiments at Kyoto University Critical Assembly

Sample reactivity and void reactivity experiments are carried out in the solid-moderated and solid-reflected cores at the Kyoto University Critical Assembly (KUCA) with the combined use of aluminum (Al), lead (Pb), and bismuth (Bi) samples, and Al spacers simulating the void. MCNP6.2 eigenvalue calculations together with JENDL-4.0 provide good accuracy of sample reactivity with the comparison of experimental results. Also, experimental void reactivity is attained by using MCNP6.2 together with JENDL-4.0 and ENDF/B-VII.1 with a small relative difference between experiments and calculations. Uncertainty in sample reactivity and void reactivity due to the ENDF/B-VII.1 Al, Pb, and Bi nuclear data is quantified using sensitivities calculated by the ksen card in MCNP6.2 and covariances provided by SCALE6.2. A series of reactivity analyses with the Al spacer simulating the void demonstrates the means of analyzing the void in the solid-moderated and solid-reflected cores at KUCA.

Measurement of 233U-HU Substitution Reactivity Worth in KUCA for Integral Validation of 233U Nuclear Data

In order to perform an integral evaluation of 233U nuclear data, measurements of substitution reactivity worth between 233U sample and high enriched uranium (HU) sample in the Kyoto University Critical Assembly (KUCA) were carried out. In this study, the experimental core was consisted of 24 fuel elements and 1 special fuel element with the 233U or HU sample. The fuel element was consisted of 31 unit cells and sandwiched by a upper and a lower polyethylene reflector. The unit has one enriched uranium plate of 1/16” thickness, three polyethylene plates of 1/8” thickness. The special fuel element was loaded at the center of the core. The substitution reactivity worth was defined as difference between the excess reactivity of the core in which the 233U sample or the HU sample inserted the core. The excess reactivities ware measured by the positive period method. As the result, the substation reactivity was obtained 0.0146±0.0006 %∆k/k. A numerical result by MVP3 with JENDL-4.0 was 0.0141±0.0002 %∆k/k and a C/E was 0.964 ± 0.041. On the other hand, the calculated substitution reactivity worth with JENDL-5 was 0.0168±0.0002 %∆k/k and a C/E was 1.151±0.047.

Measurement of double-differential neutron yields for iron, lead, and bismuth induced by 107-MeV protons for research and development of accelerator-driven systems

For the research and development of accelerator-driven systems (ADSs) and fundamental ADS reactor physics research using the Kyoto University Critical Assembly combined with the fixed-field alternating gradient (FFAG) accelerator, we are conducting a series of experiments on double-differential neutron yields using the FFAG accelerator at Kyoto University. This paper presents an overview of the experiments together with preliminary results.

Monte Carlo sensitivity analyses of isothermal temperature coefficient in solid-moderated and solid-reflected cores at Kyoto University critical assembly

The Monte Carlo sensitivity analyses are carried out for the isothermal temperature coefficient (ITC) of two solid-moderated and solid-reflected cores characterized by medium-fast and thermal spectra at the Kyoto University Critical Assembly. From the isotope-wise and region-wise sensitivity analyses, the ITCs of two cores are found to be mainly determined by variations of the thermal scattering cross section of hydrogen in the moderator and reflector regions due to the spectral shift effect. The ITC contribution of the moderator is negative due to the degradation of neutron moderation while that of the reflector is positive due to a decrease in the neutron absorption in both cores. As the 1H/235U ratio is smaller, the ITC contribution of the moderator is less negative due to the smaller degradation of neutron moderation, while that of the reflector is more positive due to the larger thermal peak of neutron flux distribution in the reflector region.

Measurement of 107-MeV proton-induced double-differential thick target neutron yields for Fe, Pb, and Bi using a fixed-field alternating gradient accelerator at Kyoto University

ABSTRACT Double-differential thick target neutron yields (TTNYs) for Fe, Pb, and Bi targets induced by 107-MeV protons were measured using the fixed-field alternating gradient accelerator at Kyoto University for the research and development of accelerator-driven systems (ADSs) and fundamental ADS reactor physics research at the Kyoto University Critical Assembly (KUCA). Note that TTNYs were obtained with the time-of-flight method using a neutron detector system comprising eight neutron detectors; each detector has a small EJ-301 liquid organic scintillator and photomultiplier tube. The TTNYs obtained were compared with calculation results using Monte Carlo-based spallation models (i.e. INCL4.6/GEM, Bertini/GEM, JQMD/GEM, and JQMD/SMM/GEM) and the evaluated high-energy nuclear data library, i.e. JENDL-4.0/HE, implemented in the particle and heavy ion transport code system (PHITS). All models, including JENDL-4.0/HE, failed to predict high-energy peaks at a detector angle of . Comparing the energy- and angle-integrated spallation neutron yields at energies of ≤20 MeV estimated using the measured TTNYs and the PHITS indicated that INCL4.6/GEM would be suitable for the Monte Carlo transport simulation of ADS reactor physics experiments at the KUCA.

Nuclear Data–Induced Uncertainty of Criticality in Solid-Moderated and Solid-Reflected Cores with Highly Enriched Uranium Fuel at Kyoto University Critical Assembly

Nuclear data–induced uncertainty of criticality is successfully analyzed by combining the eigenvalue calculations, the uncertainty, and the reduction of uncertainty with the use of the KENO-VI code, the TSUNAMI-3D and the TSURFER modules of the SCAL6.2.4 code system, respectively. The comparative study of conventional and revised S(α, β) applications is also conducted by KENO-VI. Notably, the KENO-VI analyses reveal the difference between the experimental and numerical results of criticality and the neutron spectrum dependence of criticality on the H/U ratio in the solid-moderated and solid-reflected cores at the Kyoto University Critical Assembly (KUCA). The difference is identified as the leading cause of uncertainty in the 235U fission spectrum (χ value) through the combined use of the uncertainty and the cross-section adjustment by TSUNAMI-3D and TSURFER, respectively, especially that the highly enriched uranium (HEU) fuel is loaded into the KUCA cores. Also, the neutron spectrum dependence of criticality is attributable to the uncertainty induced by the cross-section data of 235U capture, 27Al elastic scattering, and inelastic scattering reactions in the HEU fuel plate and to the 1H capture reactions in the polyethylene moderator through the TSUNAMI-3D analyses.

Application of dynamic mode decomposition to Rossi-α method in a critical state using file-by-file moving block bootstrap method

ABSTRACT Prompt neutron decay constant in a critical state is useful information to validate the numerically predicted ratio of the point kinetics parameters , where and are the effective delayed neutron fraction and prompt neutron lifetime, respectively. To directly measure in a target critical system, this study proposes the application of the dynamic mode decomposition (DMD) to the reactor noise analysis based on the Rossi- method. The DMD-based Rossi- method enables us to robustly estimate the fundamental mode component of from the Rossi- histograms measured using multiple neutron detectors. Furthermore, the file-by-file moving block bootstrap method is newly proposed for the statistical uncertainty quantification of to prevent huge memory usage when the neutron count rate is high and/or the total measurement time is long. A critical experiment has been conducted at Kyoto University Critical Assembly to demonstrate the proposed method. As a result, the proposed method can uniquely determine the value of which the statistical uncertainty is smallest. By utilizing this experimental result of , numerical results of ratio using the continuous energy Monte Carlo code MCNP6.2 with recent nuclear data libraries, which are processed by the nuclear data processing code FRENDY, are validated.

Uncertainty quantification of light-water-moderated and light-water-reflected cores with highly-enriched uranium fuel at Kyoto University Critical Assembly

ABSTRACT A series of uncertainty quantifications are successfully conducted by the eigenvalue calculations, the uncertainty of criticality and reactivity, and the reduction of uncertainty using the KENO-VI code, the TSUNAMI-3D, the TSAR and the TSURFER modules of the SCALE code system, respectively. The eigenvalue calculations with KENO-VI provide noteworthy the discrepancy between experimental and numerical results of effective multiplication factor (keff), and the neutron spectrum dependence of keff in the light-water-moderated and light-water-reflected cores (C-cores) at the Kyoto University Critical Assembly (KUCA). In terms of TSUNAMI-3D, the discrepancy of keff is determined mainly by the uncertainty of uranium-235 (235U) fission spectrum (χ value) induced by the cross-section data, due to the use of highly-enriched uranium fuel in the KUCA C-cores. Also, the neutron spectrum dependence of keff is adequately understood from the effects of the uncertainty of hydrogen-1 capture, oxygen-16 elastic scattering, and aluminum-27 elastic and inelastic scattering reactions, in addition to the uncertainty of 235U fission spectrum. The reduction in the uncertainty of criticality and reactivity is interestingly achieved by the cross-section adjustment method provided in the TSURFER module, demonstrating the contribution of relevant nuclides and reactions of core components in the KUCA C-cores.

Applicability of Dynamic Mode Decomposition to Estimate Fundamental Mode Component of Prompt Neutron Decay Constant from Experimental Data

To robustly estimate the fundamental mode component of prompt neutron decay constant α in a subcritical system, dynamic mode decomposition (DMD) is applied to time-series data obtained by the pulsed-neutron source (PNS) and Rossi-α methods. For the statistical uncertainty quantification of α by DMD, randomly sampled virtual data are used for the DMD procedure. The applicability of DMD is demonstrated by analyzing the experimental results by the PNS and Rossi-α methods, which are performed at the Kyoto University Critical Assembly (KUCA). When applying the DMD to the PNS and Rossi-α experimental data, a constant signal was added to the experimental data to remove the background constant component. The application results indicate that DMD enables one to robustly estimate the fundamental mode component of α in the PNS and Rossi-α methods.

Experimental Analyses of 243Am and 235U Fission Reaction Rates at Kyoto University Critical Assembly

Measurements of 243Am and 235U fission reaction rates are conducted with the use of two single fission chambers in the solid-moderated and solid-reflected core at the Kyoto University Critical Assembly (KUCA). Critical irradiation experiments of 243Am and 235U foils are carried out, and the measured result of 243Am/235U is 0.0424 ± 0.0019; also, calculation/experiment values between calculated (MCNP6.1 with JENDL-4.0, ENDF/B-VIII.0, and JEFF-3.3) and measured results of 243Am/235U range among 0.93 ± 0.04, 0.94 ± 0.04, and 0.93 ± 0.04, respectively. Through a comparison between the measured and calculated results, the 243Am fission cross-section data of the three major nuclear data libraries are successfully validated, demonstrating the same accuracy as that of previous minor actinide irradiation experiments at KUCA. Importantly, the comparison also provides the complemental data of integral experiments of 243Am fission reaction rates that confirm the accuracy of the 243Am fission cross-section data.

Proposal and applicability of estimated criticality lower-limit multiplication factor using the bootstrap method

ABSTRACT To judge whether an application system is a subcritical state or not based on numerical results of the effective neutron multiplication factor , an evaluation method of the estimated criticality lower-limit multiplication factor (ECLLMF) using the bootstrap method is newly proposed. By utilizing numerical results of for critical benchmark-problems that are selected depending on neutronic similarity to the application system, the ECLLMF should be carefully and conservatively estimated based on uncertainties of due to a criticality safety analysis code, experimental uncertainty, and covariance matrix of the nuclear data library. Furthermore, the frequency distribution of for these problems does not necessarily obey an ideal normal distribution. Using a resampling technique called ‘bootstrap method,’ the proposed method can reasonably estimate the ECLLMF considering the uncertainties and nuclear data-induced correlation between each critical benchmark-problem without the assumption of normality. To investigate the applicability of the proposed method, the approach-to-criticality experiment was carried out at the Kyoto University Critical Assembly (KUCA). Comparison of numerical results of and the ECLLMF using the bootstrap method indicated that the proposed method was able to judge an actual subcritical core as subcritical state.

Development of a Wide Dynamic Range Neutron Flux Measurement Instrument Having Fast Time Response for Fusion Experiments

A wide-range neutron flux measurement instrument is developed herein for monitoring the total neutron emission rate and yield of the Large Helical Device (LHD) during deuterium experiments implemented from March 2017 in the National Institute for Fusion Science (NIFS), Japan. The instrument is designed for and installed on the Neutron Flux Monitoring (NFM) system, which measures the counting rate using a 235U Fission Chamber. By combining the pulse counting and Campbell methods, the Digital Signal Processing Unit (DSPU) realized a wide dynamic range of over six orders of magnitude from 1 × 103 counts/s (cps) to 5 × 109 cps. This study explains and discusses how the instrument is developed, including topics from the predevelopment activities to the verification test at the Kyoto University Critical Assembly (KUCA). Experimental results in the LHD using the finished products suggest that the NFM system works well during deuterium experiments.

Uncertainty Quantification of Lead and Bismuth Sample Reactivity Worth at Kyoto University Critical Assembly

Uncertainty quantification of lead (Pb) and bismuth (Bi) sample reactivity worth is numerically determined using the SCALE6.2 code system and experimental results obtained from the solid-moderated and solid-reflected core at the Kyoto University Critical Assembly (KUCA) to demonstrate the sensitivity coefficients of aluminum (Al) and Bi scattering reactions. From the results of the numerical analyses, the impact of 27Al and 209Bi scattering cross sections obtained using SCALE6.2/TSAR is disclosed on the Bi sample reactivity worth using Al reference and Bi test samples, although the uncertainty itself is small in the Bi sample reactivity worth. In future studies, the actual impact of 209Bi inelastic scattering reactions in liquid Pb-Bi eutectic needs to be considered under numerical simulations of the void reactivity in the accelerator-driven system. Also, in the KUCA experiments, conventional modeling of void evolution by the Al reference sample is expected to be altered by a void (air) in terms of the 27Al elastic and inelastic scattering reactions of the Al reference sample.

REACTOR PHYSICS EXPERIMENT IN A GRAPHITE-MODERATION SYSTEM FOR HTGR

The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs). The objectives are to introduce a generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to introduce reactor noise analysis to High Temperature Engineering Test Reactor (HTTR) experiment to observe sub-criticality. To achieve the objectives, the reactor core of graphite-moderation system named B7/4”G2/8”p8EUNU+3/8”p38EU(1) was newly composed in the B-rack of Kyoto University Critical Assembly (KUCA). The core is composed of the fuel assembly, driver fuel assembly, graphite reflector, and polyethylene reflector. The fuel assembly is composed of enriched uranium plate, natural uranium plate and graphite plates to realize the average fuel enrichment of HTTR and it’s spectrum. However, driver fuel assembly is necessary to achieve the criticality with the small-sized core. The core plays a role of the reference core of the bias factor method, and the reactor noise was measured to develop the noise analysis scheme. In this study, the overview of the criticality experiments is reported. The reactor configuration with graphite moderation system is rare case in the KUCA experiments, and this experiment is expected to contribute not only for an HTGR development but also for other types of a reactor in the graphite moderation system such as a molten salt reactor development.

REACTOR NOISE ANALYSIS FOR A GRAPHITE-MODERATED AND -REFLECTED CORE IN KUCA

In graphite-reflected thermal reactors, even a detector placed far from fuel region may detect a certain degree of the correlation amplitude. This is because mean free path of neutrons in graphite is longer than that in water or polyethylene. The objective of this study is experimentally to confirm a high flexibility of neutron detector placement in graphite reflector for reactor noise analysis. The present reactor noise analysis was carried out in a graphite-moderated and -reflected thermal core in Kyoto University Critical Assembly (KUCA). BF3 proportional neutron counters (1” dia.) were placed in graphite reflector region, where the counters were separated by about 35cm and 30cm -thick graphite from the core, respectively. At a critical state and subcritical states, time-sequence signal data from these counters were acquired and analyzed by a fast Fourier transform (FFT) analyzer, to obtain power spectral density in frequency domain. The auto-power spectral density obtained from the counters far from the core contained a significant degree of correlated component. A least-squares fit of a familiar formula to the auto-power spectral density data was made to determine the prompt-neutron decay constant. The decay constant was 63.3±14.5 [1/s] in critical state. The decay constant determined from the cross-power spectral density and coherence function data between the two counters also had a consistent value. It is confirmed that reactor noise analysis is possible using a detector placed at about 35cm far from the core, as we expected.

SUBCRITICALITY ESTIMATION BY VIRTUAL NEUTRON CAPTURE METHOD

The criticality safety control technique is required for the fuel debris removal from the Fukushima Dai-Ichi Nuclear Power Station which experienced a severe accident. The subcriticality estimation is expected to be done with only limited information about the fuel debris while the primary containment vessel internal survey work is ongoing. The purpose of this study is to develop the subcriticality estimation method called the virtual neutron capture method. The neutrons from the surface of the fuel debris represent a major portion of detector counts. The method consists of two evaluations: the evaluation at the surface of the fuel debris for which the isotope compositions are known by fuel debris sampling and the evaluation at the region of the fuel debris for which these compositions are unknown. For the unknown composition region, the average isotope composition with arbitrary water content is given. The method surveys the relationship with the detector count and the neutron multiplication factor with any size of the unknown composition region and any ratio of the water content before the on-site evaluation. The method is verified by experiments done in the Kyoto University Critical Assembly. The method shows that the maximum difference from the reference neutron multiplication factor is 4.5 %dk. As a result, the virtual neutron capture method can be adopted to the subcriticality monitoring if the method includes the estimation margin of 4.5 %dk within the neutron multiplication factor range from 0.70 to 0.95.