Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 86

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Status of the uncertainty quantification for severe accident sequences of different NPP-designs in the frame of the H-2020 project MUSA

Brumm, S.*; Gabrielli, F.*; Sanchez-Espinoza, V.*; Groudev, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; Bocanegra, R.*; Herranz, L. E.*; Berda$"i$, M.*; et al.

Proceedings of 10th European Review Meeting on Severe Accident Research (ERMSAR 2022) (Internet), 13 Pages, 2022/05

Journal Articles

Quasi-Monte Carlo sampling method for simulation-based dynamic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Journal of Nuclear Science and Technology, 59(3), p.357 - 367, 2022/03

 Times Cited Count:0

Dynamic probabilistic risk assessment (PRA), which handles epistemic and aleatory uncertainties by coupling the thermal-hydraulics simulation and probabilistic sampling, enables a more realistic and detailed analysis than conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a station blackout sequence in a boiling water reactor and compared each method. The result indicated that quasi-Monte Carlo sampling method handles the uncertainties most effectively in the assumed scenario.

Journal Articles

Great achievements of M. Salvatores for nuclear data adjustment study with use of integral experiments

Yokoyama, Kenji; Ishikawa, Makoto*

Annals of Nuclear Energy, 154, p.108100_1 - 108100_11, 2021/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In the design of innovative nuclear reactors such as fast reactors, the improvement of the prediction accuracies for neutronics properties is an important task. The nuclear data adjustment is a promising methodology for this issue. The idea of the nuclear data adjustment was first proposed in 1964. Toward its practical application, however, a great deal of study has been conducted over a long time. While it took about 10 years to establish the theoretical formulation, the research and development for its practical application has been conducted for more than half a century. Researches in this field are still active, and the fact suggests that the improvement of the prediction accuracies is indispensable for the development of new types of nuclear reactors. Massimo Salvatores, who passed away in March 2020, was one of the first proposers to develop the nuclear data adjustment technique, as well as one of the great contributors to its practical application. Reviewing his long-time works in this area is almost the same as reviewing the history of the nuclear data adjustment methodology. The authors intend that this review would suggest what should be done in the future toward the next development in this area. The present review consists of two parts: a) the establishment of the nuclear data adjustment methodology and b) the achievements related to practical applications. Furthermore, the former is divided into two aspects: the study on the nuclear data adjustment theory and the numerical solution for sensitivity coefficient that is requisite for the nuclear data adjustment. The latter is separated to three categories: the use of integral experimental data, the uncertainty quantification and design target accuracy evaluation, and the promotion of nuclear data covariance development.

Journal Articles

Stochastic estimation of radionuclide composition in wastes generated at Fukushima Daiichi Nuclear Power Station using Bayesian inference

Sugiyama, Daisuke*; Nakabayashi, Ryo*; Tanaka, Shingo*; Koma, Yoshikazu; Takahatake, Yoko

Journal of Nuclear Science and Technology, 58(4), p.493 - 506, 2021/04

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Impact of soil erosion potential uncertainties on numerical simulations of the environmental fate of radiocesium in the Abukuma River basin

Ikenoue, Tsubasa; Shimadera, Hikari*; Kondo, Akira*

Journal of Environmental Radioactivity, 225, p.106452_1 - 106452_12, 2020/12

 Times Cited Count:1 Percentile:21.44(Environmental Sciences)

This study focused on the uncertainty of the factors of the Universal Soil Loss Equation (USLE) and evaluated its impacts on the environmental fate of $$^{137}$$Cs simulated by a radiocesium transport model in the Abukuma River basin. The USLE has five physically meaningful factors: the rainfall and runoff factor (R), soil erodibility factor (K), topographic factor (LS), cover and management factor (C), and support practice factor (P). The simulation results showed total suspended sediment and $$^{137}$$Cs outflows were the most sensitive to C and P among the all factors. Therefore, land cover and soil erosion prevention act have the great impact on outflow of suspended sediment and $$^{137}$$Cs. Focusing on land use, the outflow rates of $$^{137}$$Cs from the forest areas, croplands, and undisturbed paddy fields were large. This study indicates that land use, especially forest areas, croplands, and undisturbed paddy fields, has a significant impact on the environmental fate of $$^{137}$$Cs.

Journal Articles

Enhancement of the treatment of system interactions in a dynamic PRA tool

Tanaka, Yoichi; Tamaki, Hitoshi; Zheng, X.; Sugiyama, Tomoyuki

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11

Journal Articles

Total cross section model with uncertainty evaluated by KALMAN

Hashimoto, Shintaro; Sato, Tatsuhiko

EPJ Web of Conferences, 239, p.03015_1 - 03015_4, 2020/09

Particle transport simulation codes based on the Monte Carlo technique have been successfully applied to shielding calculations in accelerator facilities. Estimation of not only statistical uncertainties, which depend on the number of trials, but also systemic uncertainties, which are caused by uncertainty of total cross section models, is required to confirm the reliability of the simulation results. We evaluated unclear quantities of internal parameters included in the total cross section model by the KALMAN code, which is based on the least squares technique, comparing with experimental data of the total cross section. The uncertainties in the total cross sections obtained by the new model are comparable to the experimental errors. In the present study, the systematic uncertainty included in the simulation results can be estimated by performing the transport calculations with variation of the internal parameters within their unclear quantities.

Journal Articles

Uncertainty quantification of seismic response of reactor building considering different modeling methods

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Itoi, Tatsuya*; Takada, Tsuyoshi*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08

After the 2011 Fukushima accident, the seismic regulation for Nuclear Power Plants (NPP) have been strengthened to take countermeasures against accidents beyond design basis conditions. Therefore, the importance of seismic probabilistic risk assessment has drawn much attention. Uncertainty quantification is a very important issue in the fragility assessment for NPP buildings. In this study, the authors focus on the epistemic uncertainty that can be reduced, and aims to clarify the effects due to different modeling methods of NPP buildings on seismic response results. As the first step of this study, the authors compared the effects on seismic response using two kinds of modeling methods. In order to evaluate the effect, seismic response analysis was performed on two types of building models; the three dimensional finite element model and the conventional lumped mass with sway-rocking model. As the input ground motion, the authors adopted 200 types of simulated seismic ground motions generated by fault rupture models with stochastic seismic source characteristics. For the uncertainty quantification, the authors conducted statistical analyses of the effects on seismic response results of two kinds of modeling methods on building response for each input ground motions, and quantitatively evaluated the uncertainty of response considering different modeling methods. In particular, the difference in modeling methods clearly appeared near the openings of the floors and walls. The authors also report on the knowledge about these three-dimensional effects in seismic response analysis.

Journal Articles

Activity report of the task group of radiation protection about wastes containing natural radioactive nuclides

Saito, Tatsuo; Kobayashi, Shinichi*; Zaitsu, Tomohisa*; Shimo, Michikuni*; Fumoto, Hiromichi*

Hoken Butsuri (Internet), 55(2), p.86 - 91, 2020/06

Safety cases for disposal of uranium bearing waste and NORM with uranium has not yet been fully developed in Japan, because of safety assessment of extraordinary long timespan and uncertainty in unexpected incidents with uncompleted radon impact evaluation measures arising from uranium waste disposal facility in far future. Our task group of radiation protection for wastes with natural radioactive nuclides studied some safety cases with disposal of uranium bearing waste and NORM in terms of nuclides, U-235, U-238 and their progenies, and comprehensively discussed the current state of their disposal in comparison to the ideas of international organizations such as ICRP and IAEA. We developed our ideas for long term uncertainty and radon with the knowledge of experts in each related area of direction, repeating discussions, focusing out the orientation of each directions, and outlined the recommendations with our suggestions of solving important issues in the future to be addressed.

Journal Articles

Estimation of uncertainty in lead spallation particle multiplicity and its propagation to a neutron energy spectrum

Iwamoto, Hiroki; Meigo, Shinichiro

Journal of Nuclear Science and Technology, 57(3), p.276 - 290, 2020/03

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

This paper presents an approach to uncertainty estimation of spallation particle multiplicity of lead ($$^{rm nat}$$Pb), primarily focusing on proton-induced spallation neutron multiplicity ($$x_{pn}$$) and its propagation to a neutron energy spectrum. The $$x_{pn}$$ uncertainty is estimated from experimental proton-induced neutron-production double-differential cross sections (DDXs) and model calculations with the Particle and Heavy Ion Transport code System (PHITS). Uncertainties in multiplicities for $$(n,xn)$$, $$(p,xp)$$, and $$(n,xp)$$ reactions are then inferred from the estimated $$x_{pn}$$ uncertainty and the PHITS calculation. Using these uncertainties, uncertainty in a neutron energy spectrum produced from a thick $$^{rm nat}$$Pb target bombarded with 500 MeV proton beams, measured in a previous experiment, is quantified by a random sampling technique, and propagation to the neutron energy spectrum is examined. Relatively large uncertainty intervals (UIs) were observed outside the lower limit of the measurement range, which is prominent in the backward directions. Our findings suggest that a reliable assessment of spallation neutron energy spectra requires systematic DDX experiments for detector angles and incident energies below 100 MeV as well as neutron energy spectrum measurements at lower energies below $$sim$$1.4 MeV with an accuracy below the quantified UIs.

Journal Articles

Evaluation of the effects of differences in building models on the seismic response of a nuclear power plant structure

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Nihon Jishin Kogakkai Rombunshu (Internet), 20(2), p.2_1 - 2_16, 2020/02

AA2018-0122.pdf:2.15MB

no abstracts in English

Journal Articles

Development of calculation methodology for estimation of radionuclide composition in wastes generated at Fukushima Daiichi Nuclear Power Station

Sugiyama, Daisuke*; Nakabayashi, Ryo*; Koma, Yoshikazu; Takahatake, Yoko; Tsukamoto, Masaki*

Journal of Nuclear Science and Technology, 56(9-10), p.881 - 890, 2019/09

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

Journal Articles

Cross-section-induced uncertainty evaluation of MA sample irradiation test calculations with consideration of dosimeter data

Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto; Takeda, Toshikazu*

Annals of Nuclear Energy, 130, p.118 - 123, 2019/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In MA sample irradiation test data calculations, the neutron fluence during irradiation period is generally scaled by using dosimetry data in order to improve calculation accuracy. In such a case, appropriate correction is required to burnup sensitivity coefficients obtained by the conventional generalized perturbation theory because some cancellations occur in the burnup sensitivity coefficients. Therefore, a new formula for the burnup sensitivity coefficient has been derived with the consideration of the neutron fluence scaling effect (NFS). In addition, the cross-section-induced uncertainty is evaluated by using the obtained burnup sensitivity coefficients and the covariance data based on the JENDL-4.0.

Journal Articles

Estimation method of systematic uncertainties in Monte Carlo particle transport simulation based on analysis of variance

Hashimoto, Shintaro; Sato, Tatsuhiko

Journal of Nuclear Science and Technology, 56(4), p.345 - 354, 2019/04

 Times Cited Count:1 Percentile:23.13(Nuclear Science & Technology)

Particle transport simulations based on the Monte Carlo method have been applied to shielding calculations. Estimation of not only statistical uncertainty related to the number of trials but also systematic one induced by unclear physical quantities is required to confirm the reliability of calculated results. In this study, we applied a method based on analysis of variance to shielding calculations. We proposed random- and three-condition methods. The first one determines randomly the value of the unclear quantity, while the second one uses only three values: the default value, upper and lower limits. The systematic uncertainty can be estimated adequately by the random-condition method, though it needs the large computational cost. The three-condition method can provide almost the same estimate as the random-condition method when the effect of the variation is monotonic. We found criterion to confirm convergence of the systematic uncertainty as the number of trials increases.

Journal Articles

A Systematic radionuclide migration parameter setting approach for potential siting environments in Japan

Hamamoto, Takafumi*; Ishida, Keisuke*; Shibutani, Sanae*; Fujisaki, Kiyoshi*; Tachi, Yukio; Ishiguro, Katsuhiko*; McKinley, I. G.*

Proceedings of 2019 International High-Level Radioactive Waste Management Conference (IHLRWM 2019) (USB Flash Drive), p.77 - 82, 2019/04

Journal Articles

Improvement of plant reliability based on combining of prediction and inspection of crack growth due to intergranular stress corrosion cracking

Uchida, Shunsuke; Chimi, Yasuhiro; Kasahara, Shigeki; Hanawa, Satoshi; Okada, Hidetoshi*; Naito, Masanori*; Kojima, Masayoshi*; Kikura, Hiroshige*; Lister, D. H.*

Nuclear Engineering and Design, 341, p.112 - 123, 2019/01

 Times Cited Count:2 Percentile:42.44(Nuclear Science & Technology)

Improvement of plant reliability based on reliability-centered-maintenance (RCM) is going to be undertaken in NPPs. RCM is supported by risk-based maintenance (RBM). The combination of prediction and inspection is one of the key issues to promote RBM. Early prediction of IGSCC occurrence and its propagation should be confirmed throughout the entire plant systems which should be accomplished by inspections at the target locations followed by timely application of suitable countermeasures. From the inspections, accumulated data will be applied to confirm the accuracy of the code, to tune some uncertainties of the key data for prediction, and then, to increase their accuracy. The synergetic effects of prediction and inspection on application of effective and suitable countermeasures are expected. In the paper, the procedures for the combination of prediction and inspection are introduced.

Journal Articles

Conceptual uncertainties in modelling the interaction between engineered and natural barriers of nuclear waste repositories in crystalline rocks

Finsterle, S.*; Lanyon, B.*; ${AA}$kesson, M.*; Baxter, S.*; Bergstr$"o$m, M.*; Bockg${aa}$rd, N.*; Dershowitz, W.*; Dessirier, B.*; Frampton, A.*; Fransson, ${AA}$.*; et al.

Geological Society, London, Special Publications, No.482, p.261 - 283, 2019/00

 Times Cited Count:3 Percentile:57.07

Nuclear waste disposal in geological formations relies on a multi-barrier concept that includes engineered components which in many cases includes a bentonite buffer surrounding waste packages and the host rock. An SKB's (Swedish Nuclear Fuel and Waste Management Co.) Modelling Task Force project facilitated to improve the overall understanding of rock - bentonite interactions, as 11 teams used different conceptualisations and modelling tools to analyse the in-situ experiment at the $"A$ps$"o$ Hard Rock Laboratory. The exercise helped identify conceptual uncertainties that led to different assessments of the relative importance of the engineered and natural barrier subsystems and of aspects that need to be better understood to arrive at reliable predictions of bentonite wetting.

Journal Articles

Sensitivity and uncertainty analysis of $$beta_{rm eff}$$ for MYRRHA using a Monte Carlo technique

Iwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*

European Physical Journal; Nuclear Sciences & Technologies (Internet), 4, p.42_1 - 42_7, 2018/11

This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction $$beta_{rm eff}$$ for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo transport code MCNP. The $$beta_{rm eff}$$ sensitivities are calculated by the modified $$k$$-ratio method proposed by Chiba. Comparing the $$beta_{rm eff}$$ sensitivities obtained with different scaling factors $$a$$ introduced by Chiba shows that a value of $$a=20$$ is the most suitable for the uncertainty quantification of $$beta_{rm eff}$$. Using the calculated $$beta_{rm eff}$$ sensitivities and the JENDL-4.0u covariance data, the $$beta_{rm eff}$$ uncertainties for the critical and subcritical cores are determined to be 2.2 $$pm$$ 0.2% and 2.0 $$pm$$ 0.2%, respectively, which are dominated by delayed neutron yield of $$^{239}$$Pu and $$^{238}$$U.

Journal Articles

Application of Bayesian optimal experimental design to reduce parameter uncertainty in the fracture boundary of a fuel cladding tube under LOCA conditions

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

86 (Records 1-20 displayed on this page)