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Journal Articles

Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03

Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by $$^{6}$$Li(n$$_{t}$$,$$alpha$$)$$^{3}$$H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.

Journal Articles

Evaluation of tritium release into primary coolant for research and testing reactors

Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of $$^{9}$$Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a research and testing reactors containing beryllium reflectors.

Journal Articles

Thick target neutron yields from 100- and 230-MeV/nucleon helium ions bombarding water, PMMA, and iron

Tsai, P.-E.; Heilbronn, L. H.*; Lai, B.-L.*; Iwata, Yoshiyuki*; Murakami, Takeshi*; Sheu, R.-J.*

Nuclear Instruments and Methods in Physics Research B, 449, p.62 - 70, 2019/06

 Times Cited Count:0 Percentile:100(Instruments & Instrumentation)

The secondary neutrons produced from 100- and 230-MeV/nucleon He ions, respectively, stopping in the thick iron, PMMA and water targets are measured by the time of flight method combined with the pulse shape discrimination of liquid scintillators. The experimental data show that the secondary neutrons were contributed from breakup of projectiles, emission from overlap regions of projectile nuclei and target nuclei, and evaporation of projectiles and target nuclei. The measured double-differential thick target neutron yields, angular distributions, and total neutron yields per ion, were benchmarked by the PHITS, FLUKA, and MCNP model calculations. The default models in these codes agree the experimental data well at intermediate-to-large angles in the low-to-intermediate energy range. However, the physics models implemented in PHITS need further improvement for some particular nuclear interaction mechanisms, and this work can be an importance reference for future model development.

Journal Articles

Computation speeds and memory requirements of mesh-type ICRP reference computational phantoms in Geant4, MCNP6, and PHITS

Yeom, Y. S.*; Han, M. C.*; Choi, C.*; Han, H.*; Shin, B.*; Furuta, Takuya; Kim, C. H.*

Health Physics, 116(5), p.664 - 676, 2019/05

 Times Cited Count:3 Percentile:20.8(Environmental Sciences)

Recently, Task Group 103 of the ICRP developed the mesh-type reference computational phantoms (MCRPs), which are planned for use in future ICRP dose coefficient calculation. Performance of major Monte Carlo particle transport codes (Geant4, MCNP6, and PHITS) were tested with MCRP. External and internal exposure of various particles and energies were calculated and the computational times and required memories were compared. Additionally calculation for voxel-mesh phantom was also conducted so that the influence of different mesh-representation in each code was studied. Memory usage of MRCP was as large as 10 GB with Geant4 and MCNP6 while it is much less with PHITS (1.2 GB). In addition, the computational time required for MRCP tends to increase compared to voxel-mesh phantoms with Geant4 and MCNP6 while it is equal or tends to decrease with PHITS.

Journal Articles

Implementation of random sampling for ACE-format cross sections using FRENDY and application to uncertainty reduction

Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi

Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.1493 - 1502, 2019/00

A perturbation capability of ACE formatted cross section files was developed using the modules of FRENDY. Uncertainty quantification using MCNP was carried out for the Godiva critical experiment by the RS method. We verified the results of the RS method by comparing with those obtained by the conventional sensitivity analyses. Moreover, uncertainty reduction using the bias factor method with the RS technique was applied to kinetic parameter, i.e., neutron generation time.

JAEA Reports

Calculations of Tritium Recoil Release from Li and U Impurities in Neutron Reflectors (Joint research)

Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2018-010, 33 Pages, 2018/11

JAEA-Technology-2018-010.pdf:2.58MB

As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.

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

Monte Carlo uncertainty quantification of the effective delayed neutron fraction

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

Journal of Nuclear Science and Technology, 55(5), p.539 - 547, 2018/05

 Times Cited Count:6 Percentile:15.83(Nuclear Science & Technology)

The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction $$beta_{rm eff}$$ is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of $$beta_{rm eff}$$ uncertainty. For validation of the calculated $$beta_{rm eff}$$, a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified $$k$$-ratio method is superior to Bretscher's prompt $$k$$-ratio method in terms of reducing the statistical uncertainty in calculating not only $$beta_{rm eff}$$ but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified $$k$$-ratio method is the most practical for uncertainty quantification of $$beta_{rm eff}$$. Finally, total $$beta_{rm eff}$$ uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of $$^{235}$$U.

Journal Articles

Multi-threading performance of Geant4, MCNP6, and PHITS Monte Carlo codes for tetrahedral-mesh geometry

Han, M. C.*; Yeom, Y. S.*; Lee, H. S.*; Shin, B.*; Kim, C. H.*; Furuta, Takuya

Physics in Medicine & Biology, 63(9), p.09NT02_1 - 09NT02_9, 2018/05

 Times Cited Count:2 Percentile:68.35(Engineering, Biomedical)

The multi-threading computation performances of the Geant4, MCNP6, and PHITS codes were evaluated using three tetrahedral-mesh phantoms with different complexity. Photon and neutron transport simulations were conducted and the initialization time, calculation time, and memory usage were measured as a function of the number of threads N used in the simulation. The initialization time significantly increases with the complexity of the phantom, but not much with the number of the threads. For the calculation time, Geant4 showed good parallelization efficiency with multi-thread computation (30 times speed-up factor for N = 40) adopting the private tallies while saturation of the speed-up factor were observed in MCNP6 and PHITS (10 and a few times for N = 40) due to the time delay for the sharing tallies. On the other hand, Geant4 requires larger memory specification and the memory usage rapidly increases with the number of threads compared to MCNP6 or PHITS. It is notable that when compared to the other codes, the memory usage of PHITS is much smaller, regardless of both the complexity of the phantom and the number of the threads.

Journal Articles

Investigation of uncertainty caused by random arrangement of coated fuel particles in HTTR criticality calculations

Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji

Annals of Nuclear Energy, 112, p.42 - 47, 2018/02

 Times Cited Count:3 Percentile:40.19(Nuclear Science & Technology)

Journal Articles

Benchmark of neutron production cross sections with Monte Carlo codes

Tsai, P.-E.; Lai, B.-L.*; Heilbronn, L. H.*; Sheu, R.-J.*

Nuclear Instruments and Methods in Physics Research B, 416, p.16 - 29, 2018/02

 Times Cited Count:1 Percentile:75.54(Instruments & Instrumentation)

Fifteen thin target experiments were selected for this benchmark study of the neutron production cross sections. The studied cases include a mix combination of $$^{12}$$C, $$^{20}$$Ne $$^{40}$$Ar, $$^{84}$$Kr, and $$^{132}$$Xe ions bombarding $$^{nat}$$Li, $$^{nat}$$C, $$^{nat}$$Al, $$^{nat}$$Cu, and $$^{nat}$$Pb target with projectile energies between 135 and 600 MeV/nucleon. The experimental data were compared to the model calculations performed by (1) PHITS version 2.73 with JQMD and GEM model, (2) PHITS version 2.82 with revised JQMD 2.0 and GEM model, (3) FLUKA version 2011.2c with RQMD 2.4 and FLUKA's own de-excitation model, and (4) MCNP6 version 1.0 with LAQGSM 03.03 and GEM2 model. This study provides useful information not only for code users but also for model developers, and it will lead to future improvements of the PHITS - JQMD model for heavy-ion induced reactions, which is important for accelerator facilities, heavy-ion radiotherapy, and space radiation.

Journal Articles

FRENDY; A New nuclear date processing system being developed at JAEA

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

EPJ Web of Conferences, 146, p.02028_1 - 02028_5, 2017/09

 Times Cited Count:0 Percentile:100

JAEA has started to develop new nuclear data processing system FRENDY (FRom Evaluated Nuclear Data libralY to any application). In this presentation, the outline of the development of FRENDY is presented. And functions and performances of FRENDY are demonstrated by generation and validation of the continuous energy cross section data libraries for MVP, PHITS and MCNP codes.

Journal Articles

Development and verification of a new nuclear data processing system FRENDY

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

Journal of Nuclear Science and Technology, 54(7), p.806 - 817, 2017/07

AA2016-0417.pdf:1.93MB

 Times Cited Count:9 Percentile:13.35(Nuclear Science & Technology)

JAEA has developed an evaluated nuclear data library JENDL and several nuclear analysis codes such as MARBLE2, SRAC, MVP and PHITS. Though JENDL and these computer codes have been widely used in many countries, the nuclear data processing system to generate the data library for application programs had not been developed in Japan and foreign nuclear data processing systems, e.g., NJOY and PREPRO are used. To process the new library for JAEA's computer codes immediately and independently, JAEA started to develop the new nuclear data processing system FRENDY in 2013. In this paper, outline, function, and verification of FRENDY are described.

Journal Articles

Feasibility study of advanced measurement technology for solution monitoring at reprocessing plant; Dose rate measurement for the solution including Pu with FP

Matsuki, Takuya; Yamanaka, Atsushi; Sekine, Megumi; Suzuki, Satoshi*; Yasuda, Takeshi; Tsutagi, Koichi; Tomikawa, Hirofumi; Nakamura, Hironobu; LaFleur, A. M.*; Browne, M. C.*

Proceedings of INMM 58th Annual Meeting (Internet), 8 Pages, 2017/07

The Tokai Reprocessing Plant (TRP) has been developing a new detector from 2015 to 2017 for purpose to monitor Pu amount in High Active Liquid Waste (HALW) containing FP. It can make a contribution to an advanced approach to effectively and efficiently conduct safeguards for reprocessing facilities because it becomes available to monitor and verify nuclear material movement continuously by a new detector, which has proposed by IAEA. For the second step of this project, we conducted dose rate measurement on the guide rail installing in the cell storing the HALW tank and comparison between measured dose rate distribution and calculation result by MCNP simulation in order to investigate the dose rate distribution which is needed for shielding design of a new detector that is used for radiation (neutron/$$gamma$$ spectrum) measurement in the cell and inquest on the monitoring position of the detector for Pu monitoring. In this paper, we report the result of the dose rate measurement in the cell, improvement of the simulation model which is cleared by comparison between measurement result and calculation result and our future plan.

Journal Articles

Numerical investigation of the random arrangement effect of coated fuel particles on the criticality of HTTR fuel compact using MCNP6

Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji

Annals of Nuclear Energy, 103, p.114 - 121, 2017/05

 Times Cited Count:5 Percentile:31.96(Nuclear Science & Technology)

Journal Articles

Benchmark study on realized random packing model for coated fuel particles of HTTR using MCNP6

Ho, H. Q.; Morita, Keisuke*; Honda, Yuki; Fujimoto, Nozomu*; Takada, Shoji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

Journal Articles

Integral test of international reactor dosimetry and fusion file on graphite assembly with DT neutron at JAEA/FNS

Ota, Masayuki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara

Fusion Engineering and Design, 98-99, p.1847 - 1850, 2015/10

 Times Cited Count:2 Percentile:74.39(Nuclear Science & Technology)

International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), has been released from the International Atomic Energy Agency (IAEA) recently. In order to validate and test IRDFF 1.0, IAEA has initiated a new Co-ordinated Research Project (CRP). Under this CRP, we have performed an integral experiment on a graphite pseudo-cylindrical slab assembly with DT neutron source at JAEA/FNS. The graphite assembly of 31.4 cm in equivalent radius and 61 cm in thickness is placed at a distance of about 20 cm from the DT neutron source. A lot of foils for the dosimetry reactions in IRDFF1.0 are inserted into the small spaces between the graphite blocks along the center axis of the assembly. After DT neutron irradiation, reaction rates for the dosimetry reactions are measured by the foil activation technique. This experiment is analyzed by using Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0. The experimental assembly and DT neutron source are modeled precisely in the MCNP calculation. The reaction rates calculated with IRDFF 1.0 as the response functions for the dosimetry reactions are compared with the experimental values. Also the calculations with JENDL Dosimetry File 99 (JENDL/D-99) are performed for comparison. The results calculated with IRDFF 1.0 show good agreement with the experimental results.

Journal Articles

Analyses of iron and concrete shielding experiments at JAEA/TIARA with JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0

Konno, Chikara; Ochiai, Kentaro; Sato, Satoshi; Ota, Masayuki

Fusion Engineering and Design, 98-99, p.2178 - 2181, 2015/10

 Times Cited Count:4 Percentile:54.36(Nuclear Science & Technology)

We have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which were supplied from JAEA, BNL and IAEA, respectively, were used for this analysis. The followings are found out from the results. (1) The calculations with JENDL/HE-2007 agree with all the measured ones well; (2) Those with ENDF/B-VII.1 tend to overestimate the measured ones with the thickness of the assemblies largely; (3) Those with FENDL-3.0 agree with the measured ones well for the iron experiment, while they overestimate the measured ones well for the concrete experiment largely. Some data in ENDF/B-VII.1 and FENDL-3.0 should be revised.

Journal Articles

Inter-code comparison of TRIPOLI${textregistered}$ and MVP on the MCNP criticality validation suite

Brun, E.*; Zoia, A.*; Trama, J. C.*; Lahaye, S.*; Nagaya, Yasunobu

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.351 - 360, 2015/09

This paper presents a joint work conducted at CEA Saclay and JAEA Tokai aimed at comparing the Monte Carlo codes TRIPOLI${textregistered}$ and MVP on a selection of ICSBEP benchmarks. Our goal is to establish a common set of Monte Carlo input decks, as a basis for rigorous inter-code comparison in criticality-safety. As a reference, we will use the MCNP Criticality Validation Suite: other Monte Carlo developers might easily join that effort in the future. For the purpose of inter-code comparison, the TRIPOLI${textregistered}$ and MVP input decks have been translated from those of MCNP, without any further assumptions. Both TRIPOLI${textregistered}$ and MVP have been run with the same ENDF/B-VII.0 evaluated nuclear data, and as far as possible the same simulation options as in the original LANL work. In this abstract, we present preliminary results on the BIGTEN benchmark. In the full paper these will be extended to the 31 benchmarks of the MCNP Criticality Validation Suite. In the future, this database will also help in the analysis of sensitivity to nuclear data, CPU times and figures of merit.

JAEA Reports

SWAT4.0; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*

JAEA-Data/Code 2014-028, 152 Pages, 2015/03

JAEA-Data-Code-2014-028.pdf:13.39MB

There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0.

235 (Records 1-20 displayed on this page)