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Ho, H. Q.; 石塚 悦男; 飯垣 和彦
Recent Contributions to Physics, 82(3), p.16 - 20, 2022/09
Detailed neutron flux distribution is important to understand the neutronic behavior during operation as well as to precise the core optimization and safety analysis of a reactor. In the literature, no calculations have been performed to show the detailed neutron flux map for the high temperature engineering test reactor (HTTR) because of the limitation of the old neutronic codes and the low performance of the computing system. The present work deals with MCNP6 Monte-Carlo calculation to determine the detailed neutron flux map in the HTTR during normal operation. At first, the calculation of neutron flux at several positions in the reactor was validated by comparing the corresponding reaction rate between the calculation and measurement. After that detailed neutron flux with the small cells of 1cm 
1cm 10cm was obtained for the entire reactor core using the fmesh tally of MCNP6 code.
藤本 望*; 多田 健一; Ho, H. Q.; 濱本 真平; 長住 達; 石塚 悦男
Annals of Nuclear Energy, 158, p.108270_1 - 108270_8, 2021/08
被引用回数：3 パーセンタイル：45.99(Nuclear Science & Technology)Japan Atomic Energy Agency has developed a new nuclear data processing code, namely FRENDY, to generate the ACE files from various nuclear libraries. A code-to-experiment verification of FRENDY processing was carried out in this study with criticality benchmark assessments of the high temperature engineering test reactor. The ACE files of the JENDL-4.0 and ENDF-B-VII.1 was generated successfully by FRENDY. These ACE files have been used in MCNP6 transportation calculation for various benchmark problems of the high temperature engineering test reactor. As a result, the k
and reaction rate obtained by MCNP6 calculation presented a good agreement compared to the experimental data. The proper ACE files generation by FRENDY was confirmed for the HTTR criticality calculations.
Yeom, Y. S.*; Han, M. C.*; Choi, C.*; Han, H.*; Shin, B.*; 古田 琢哉; Kim, C. H.*
Health Physics, 116(5), p.664 - 676, 2019/05
被引用回数：7 パーセンタイル：61.94(Environmental Sciences)国際放射線防護委員会(ICRP)のタスクグループ103により、メッシュ形状の線量評価用人体ファントム(MRCPs)の開発が進められている。この人体ファントムは、将来的には線量評価で用いる標準人体モデルとして採用される予定である。そこで、このMRCPファントムに対するベンチマーク計算を主なモンテカルロ粒子輸送計算コード(Geant4, MCNP6およびPHITS)で行った。様々な粒子およびエネルギーで外部および内部被ばくの計算を実施し、計算時間やメモリ使用量をコード間で比較した。また、ボクセルファントムに対する計算も行い、コード毎の異なるメッシュ表現による性能の違いについて調べた。MRCPのメモリ使用量はGeant4およびMCNP6で10GB程度であったのに対し、PHITSでは1.2GBと顕著に少なかった。また、計算時間に関してもGeant4およびMCNP6ではボクセルファントムに比べてMRCPの計算時間は長くなる傾向を示したが、PHITSでは同程度もしくは短縮する傾向を示した。
石塚 悦男; Kenzhina, I.*; 奥村 啓介; Ho, H. Q.; 竹本 紀之; Chikhray, Y.*
JAEA-Technology 2018-010, 33 Pages, 2018/11
JAEA-Technology-2018-010.pdf:2.58MB
試験研究炉の一次冷却材へのトリチウム放出機構解明の一環として、PHITSを用いてベリリウム、アルミニウム及び黒鉛製中性子反射体中のLi及びU不純物から反跳放出するトリチウムについて計算した。また、この結果を用いて、具体的にJMTR及びJRR-3Mのベリリウム中性子反射体を想定し、MCNP6及びORIGEN2でLi及びU不純物から生成するトリチウム量を計算してトリチウムの反跳放出量を評価した結果、Li及びU不純物から反跳放出するトリチウムは、ベリリウムから反跳放出するトリチウムに対して無視できる程度であり、それぞれ2桁及び5桁程度小さいことが明らかとなった。
Han, M. C.*; Yeom, Y. S.*; Lee, H. S.*; Shin, B.*; Kim, C. H.*; 古田 琢哉
Physics in Medicine & Biology, 63(9), p.09NT02_1 - 09NT02_9, 2018/05
被引用回数：7 パーセンタイル：39.6(Engineering, Biomedical)輸送計算コードGeant4, MCNP6, PHITSのマルチスレッド並列計算の実行性能について、異なる複雑さを持つ三体の四面体メッシュファントムを用いて調査した。ここでは、光子と中性子の輸送計算を実行し、初期化にかかる時間、輸送計算の時間及び、メモリ使用量と並列スレッド数の増加に対する相関関係を評価した。初期化にかかる時間は、ファントムの複雑化に伴い増加するものの、並列スレッド数にはあまり依存しないという傾向が三つ全ての計算コードで見られた。輸送計算の時間については、マルチスレッド並列計算に独立タリーの設計を採用しているGeant4では高い並列化効率(40並列で30倍の高速化)が見られたのに対し、MCNP6及びPHITSではタリー共有化による遅延のために、並列スレッド数増加に対する高速化の頭落ちが見られた(40並列でもMCNPは10倍、PHITSは数倍の高速化)。その一方で、Geant4は計算に必要なメモリ容量が大きく、並列スレッド数増加に対するメモリ使用量の増加もMCNP6やPHITSに比べて大きいことが分かった。また、PHITSの特筆すべき点として、メモリ使用量はファントムの複雑さやスレッド数によらず、他の二つの計算コードに比べて、顕著に小さいことも分かった。
Ho, H. Q.; 本多 友貴; 後藤 実; 高田 昌二
Annals of Nuclear Energy, 112, p.42 - 47, 2018/02
被引用回数：8 パーセンタイル：62.99(Nuclear Science & Technology)Coated fuel particle (CFP) is one of important factors attributing to the inherent safety feature of high temperature engineering test reactor (HTTR). However, the random arrangement of CFPs makes the simulation more complicated, becoming one of the factors affects the accuracy of the HTTR criticality calculations. In this study, an explicit random model for CFPs arrangement, namely realized random packing (RRP), was developed for the whole core of HTTR using a Monte-Carlo MCNP6 code. The effect of random placement of CFPs was investigated by making a comparison between the RRP and conventional uniform models. The results showed that the RRP model gave a lower excess reactivity than that of the uniform model, and the more number of fuel columns loading into the core, the greater the difference in excess reactivity between the RRP and uniform models. For example, the difference in excess reactivity increased from 0.07 to 0.17%
k/k when the number of fuel column increased from 9 to 30. Regarding the control rods position prediction, the RRP showed the results, which were closer to experiment than the uniform model. In addition, the difference in control rods position between the RRP and uniform models also increases from 12 to 17 mm as increasing number of fuel columns from 19 to 30.
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
被引用回数：3 パーセンタイル：30.05(Instruments & Instrumentation)薄膜ターゲットを用いた15種類の照射条件について、中性子生成断面積のベンチマークを実施した。条件は、核子当たり135
600MeVの
C, 
Ne, 
Ar, 
Kr, 
Xeイオン照射、
Li, C, Al, Cu, Pbターゲットを組み合わせたものである。実験値を4つのシミュレーション法(1)PHITS 2.73 (JQMDとGEMモデル)、(2)PHITS 2.82 (JQMD 2.0とGEMモデル)、(3)FLUKA 2011.2c (RQMD 2.4とFLUKA脱励起モデル)、及び(4)MCNP6-1.0 (LAQGSM 03.03とGEM2モデル)による結果と比較した。本研究は計算コードユーザーだけでなく、PHITS-JQMDモデル開発者にとっても有益な情報をもたらすものであり、加速器施設安全や重粒子線治療、宇宙放射線科学の発展に資する将来の重イオン核反応物理モデルの改良へも寄与するものである。
Ho, H. Q.; 本多 友貴; 後藤 実; 高田 昌二
Annals of Nuclear Energy, 103, p.114 - 121, 2017/05
被引用回数：8 パーセンタイル：61.27(Nuclear Science & Technology)This study investigated the random arrangement effect of Coated Fuel Particle (CFP) on criticality of the fuel compact of High-Temperature engineering Test Reactor (HTTR). A utility program coupling with MCNP6, namely Realized Random Packing (RRP), was developed to model a random arrangement of the CFPs explicitly for the specified fuel compact of HTTR. The criticality and neutronic calculations for pin cell model were performed by using the Monte Carlo MCNP6 code with an ENDF/B-VII.1 neutron library data. First, the reliability of the RRP model was confirmed by an insignificant variance of the infinite multiplication factor (k
) among 10 differently random arrangements of the CFPs. Next, the criticality of RRP model was compared with those of Non-truncated Uniform Packing (NUP) model and On-the-fly Random Packing (ORP) model which is a stochastic geometry capability in MCNP6. The results indicated that there was no substantial difference between the NUP and ORP models. However, the RRP model presented a lower k of about 0.32-0.52%k/k than the NUP model. In additions, the difference of k could be increased as the uranium enrichment decreases. The investigation of the 4-factor formula showed that the difference of k was predominantly given by the resonance escape probability, with the RRP model showing the smallest value.
Ho, H. Q.; 守田 圭介*; 本多 友貴; 藤本 望*; 高田 昌二
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
The Coated Fuel Particle plays an important role in the excellent safety feature of the High Temperature Gas-cooled Reactor. However, the random distribution of CFPs also makes the simulation of HTGR fuel become more complicated. The Monte Carlo N-particle (MCNP) code is one of the most well-known codes for validation of nuclear systems; unfortunately, it does not provide an appropriate function to model a statistical geometry explicitly. In order to deal with the stochastic media, a utility program for the random model, namely Realized Random Packing (RRP), has been developed particularly for High Temperature engineering Test Reactor (HTTR). This utility program creates a number of random points in an annular geometry. Then, these random points will be used as the center coordinate of CFPs in the MCNP6 input file and therefore the actual random arrangement of CFPs can be simulated explicitly. First, a pin-cell calculation was carried out to validate the RRP by comparing with Statistical Geometry (STG) model of MVP code. After that, the comparison between the RRP model (MCNP) and STG model (MVP) was shown in whole core criticality calculation, not only for the annular core but also for the fully-loaded core. The comparison of numerical results showed that the RRP model and STG model differed insignificantly in the multiplication factor as expected, regardless of the pin-cell or whole core calculations. In addition, the RRP model did not make the calculation time increase a lot in comparison with the conventional regular model (uniform arrangement).
Ho, H. Q.; 本多 友貴; 後藤 実; 高田 昌二; 石塚 悦男
no journal, ,
The Monte-Carlo MCNP code does not provide an appropriate model to simulate random arrangement of coated fuel particles (CFPs) in the fuel compact of high temperature engineering test reactor (HTTR). This study developed a MCNP model for the HTTR by using an explicit random method, namely realized random packing (RRP), to improve the accuracy of the benchmark assessment. Criticality results showed that by using the RRP model the accuracy of HTTR benchmark could be improved in comparison with the conventional uniform model.
多田 健一; 池原 正; 小野 道隆*; 東條 匡志*
no journal, ,
熱中性子散乱則の処理に注目し、FRENDYとNJOYの処理手法の違いについて調査し、NJOYに以下の問題点があることを特定した。(1)入力作成において高い専門性が必要で、公式なACEファイルでも適切に処理されていない場合がある、(2)作成されたACEファイルにはTHERMRで生成された全データが含まれている訳ではなく、入力した上限のエネルギー点未満のデータしか含まれない、(3)MCNP6.1では、iwt=2(ACERでの二次エネルギー分布に関するオプション)を用いて処理したACEファイルを用いると、計算が途中終了してしまう場合がある本発表ではこれらの問題点の概要とその解決策について説明する。
Ho, H. Q.; 濱本 真平; 藤本 望*; 長住 達; 後藤 実; 石塚 悦男
no journal, ,
Calculation of the decay gamma heat distribution in the graphite moderator of the HTTR is important because it affects the recriticality of the reactor during LOFC test. In another respect, the auxiliary Cf-252 neutron source of the HTTR must be replaced after a few years. So, a secondary neutron source, such as neutron emitted from (
,n) reaction on beryllium target, is considered for replacement of the Cf-252 source. For the two problems above, it is necessary to calculate the decay gamma distribution in the reactor after reactor shutdown. In this study, the decay gamma spectrum in the HTTR block was calculated by using MCNP6 and ORIGEN. The gamma spectrum from ORIGEN will become a source term in further MCNP calculation to calculate the gamma distribution in the fuel block of the HTTR.
