Impact of uncertainty reduction on lead-bismuth coolant in accelerator-driven system using sample reactivity experiments

Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro; Pyeon, C. H.*; Yamamoto, Akio*; Endo, Tomohiro*

Nuclear Science and Engineering, 198(6), p.1215 - 1234, 2024/06

https://doi.org/10.1080/00295639.2023.2246779

In this study, we have demonstrated that data assimilation 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 derived from inelastic-scattering cross-sections of lead and bismuth. We re-evaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the data assimilation formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties, and performed data assimilation 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 accelerator-driven system 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.

Measurements of the neutron total and capture cross sections and derivation of the resonance parameters of Ta

Endo, Shunsuke; Kimura, Atsushi; Nakamura, Shoji; Iwamoto, Osamu; Iwamoto, Nobuyuki; Rovira Leveroni, G.; Toh, Yosuke; Segawa, Mariko; Maeda, Makoto

Nuclear Science and Engineering, 198(4), p.786 - 803, 2024/04

https://doi.org/10.1080/00295639.2023.2227826

Experiments on criticality and reactivity worths in the FCA-XXII-1 assembly simulating highly enriched MOX fueled tight lattice LWR cores

Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu

Nuclear Science and Engineering, 24 Pages, 2024/00

https://doi.org/10.1080/00295639.2024.2347706

A series of integral experiments were conducted at FCA of JAEA, simulating LWR cores with a tight lattice cell of highly enriched MOX fuel containing more than 15% fissile plutonium. The three experimental configurations were constructed using foamed polystyrene with different void fractions to clarify the prediction accuracy of neutronic calculation codes and nuclear data among various neutron spectra. The nuclear characteristics measured in the experiments were criticality, moderator void reactivity worths, and sample reactivity worths. The preliminary analyses on experiments were conducted using a deterministic calculation code conventionally used for fast reactors with JENDL-4.0. Most reactivity worth calculations correlated well with the experimental values. Specifically for the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations were performed with MVP3 code by modeling the experimental setup in detail, confirming the validity of the deterministic calculations.

Development of experimental core configurations to clarify k variations by nonuniform core configurations

Gunji, Satoshi; Araki, Shohei; Suyama, Kenya

Nuclear Science and Engineering, 197(8), p.2017 - 2029, 2023/08

https://doi.org/10.1080/00295639.2022.2164151

The fuel debris generated by the accident at the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Plant is expected to have not only heterogeneous but also nonuniform compositions. Similarly, damaged fuel assemblies remaining in the reactor vessels also have nonuniform configurations due to some missing fuel rods. This non-uniformity may cause changing neutron multiplication factors. The effect of non-uniformity on the neutron multiplication factor is clarified by computations, and the possibility of experimentally validating the computations used for criticality management is being investigated. For this purpose, in this study the criticality effects of several core configurations of a new critical assembly, STACY, of the Japan Atomic Energy Agency with nonuniform arrangements of uranium oxide fuel rods, concrete rods, and stainless-steel rods were studied to confirm benchmarking potential. The difference in these arrangements changed the neutron multiplication factor by more than 1 $. We confirmed that changes in local neutron moderation conditions and the clustering of specific components caused this effect. In addition, the feasibility of benchmark experimental cores with nonuniform arrangements is evaluated. If benchmarking of such experiments could be realized, it would help to validate calculation codes and to develop criticality management methods by machine learning.

A Functional expansion tally method with numerical basis sets generated by singular value decomposition for one-dimensional Monte Carlo calculations

Kondo, Ryoichi; Nagaya, Yasunobu

Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2023) (Internet), 10 Pages, 2023/08

A functional expansion tally (FET) method with numerical basis functions generated by singular value decomposition (SVD) is newly proposed. Traditionally, analytical functions were used for the FET calculations, e.g., Legendre polynomials for a one-dimensional distribution. However, the expansion terms could increase to reconstruct steep or complex distributions with these functions. A basis set that can well represent the target distribution with lower order expansion is desired to achieve high accuracy with the small computational resource. In the present study, a numerical basis set is generated from snapshot data by using SVD. This approach is based on the reduced order modeling (ROM). We applied ROM to the FET method in Monte Carlo calculations. The numerical result showed the applicability of the proposed method, on the other hand, some issues were revealed, e.g., discretization of the snapshot data.

Investigation on applicability of subchannel analysis code ASFRE to thermal hydraulics analysis in fuel assembly with inner duct structure of sodium cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki

Journal of Nuclear Engineering and Radiation Science, 9(3), p.031401_1 - 031401_11, 2023/07

https://doi.org/10.1115/1.4056463

In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) investigated in JAEA, the use of a specific fuel assembly with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Since the fuel rods have an asymmetric layout by the inner duct, the validity confirmation of the numerical results of an in-house subchannel analysis code named ASFRE was required. In this paper, therefore, the code-to-code comparisons was applied with numerical results of ASFRE and those of an in-house CFD code named SPIRAL. The applicability of ASFRE was indicated through the confirmation of the consistency of specific temperature distributions.

Validation of feedback reactivity evaluation models for plant dynamics analysis code during unprotected loss of heat sink event in sodium-cooled fast reactors

Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04

https://doi.org/10.1115/1.4055328

Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.

Effects of vent size and wind on dispersion of hydrogen leaked in a partially open space; Studies by numerical analysis

Terada, Atsuhiko; Nagaishi, Ryuji

Nuclear Science and Engineering, 197(4), p.647 - 659, 2023/04

https://doi.org/10.1080/00295639.2022.2126689

In order to understand dispersion of hydrogen leaked in a partially open space practically, which can be considered as a basic model for all processes of transfer, treatment, storage and disposal of radioactive materials containing fuel debris in the decommissioning of nuclear facilities after a severe accident, by using a CFD code, the effects of vent size and outer wind on the H dispersion were analytically studied by adopting the experimental Hallway model, which has H release hole on the ceiling, one vent on the Roof vent and Door vent. Air flowed in the model from the Door vent, while H was discharged outside from the Roof vent. The discharged amount of H increased in conjunction with the air inflow when the size of Roof and or Door vents was increased. The effect of wind depended on the direction to the Door vent: wind from the same direction as the Door vent promoted the H discharge, while wind from the opposite direction suppressed. The dispersion characteristics of indoor leaked H was clarified for comparing model tests with the same Froude number and different scales. It was found from the analysis results of comparing model tests with the same Froude number and different scales that when the H leaked into the room and diffused to the air, the flow generated by the buoyancy of mixed gas created the stack effect which caused the natural ventilation by drawing in the air from the outside through vent. In addition, it was speculated that the H concentration decreased after its leak by quickly mixing with the air which flowed in from the vents and reached to the floor due to the Coanda effect, which is the effect of the free jet being drawn to a nearby wall.

Sensitivity coefficient evaluation of an accelerator-driven system using ROM-Lasso method

Katano, Ryota; Yamamoto, Akio*; Endo, Tomohiro*

Nuclear Science and Engineering, 196(10), p.1194 - 1208, 2022/10

https://doi.org/10.1080/00295639.2022.2067447

In this study, we propose the ROM-Lasso method that enables efficient evaluation of sensitivity coefficients of neutronics parameters to cross-sections. In the proposed method, a vector of sensitivity coefficients is expanded by subspace bases, so-called Active Subspace (AS) based on the idea of Reduced Order Modeling (ROM). Then, the expansion coefficients are evaluated by the Lasso linear regression between cross-sections and neutronics parameters obtained by the random sampling. The proposed method can be applied in the case where the adjoint method is difficult to be applied since the proposed method uses only forward calculations. In addition, AS is an effective subspace that can expand the vector of sensitivity coefficients with the lower number of dimension. Thus, the number of unknows is reduced from the original number of input parameters and the calculation cost is dramatically improved compared to the Lasso regression without AS. In this paper, we conducted ADS burnup calculations as a verification. We have shown how AS bases are obtained and the applicability of the proposed method.

Post-test analyses of the CMMR-4 test

Yamashita, Takuya; Madokoro, Hiroshi; Sato, Ikken

Journal of Nuclear Engineering and Radiation Science, 8(2), p.021701_1 - 021701_13, 2022/04

https://doi.org/10.1115/1.4051443