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論文

Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor

Ho, H. Q.; 石井 俊晃; 長住 達; 小野 正人; 島崎 洋祐; 石塚 悦男; 澤畑 洋明; 後藤 実; Simanullang, I. L.*; 藤本 望*; et al.

Nuclear Engineering and Design, 417, p.112795_1 - 112795_6, 2024/02

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor, $$^{252}$$Cf is used as the external neutron source. However, the $$^{252}$$Cf sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of $$^{252}$$Cf sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the $$^{252}$$Cf neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains significant even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.

論文

Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code

Simanullang, I. L.*; 中川 直樹*; Ho, H. Q.; 長住 達; 石塚 悦男; 飯垣 和彦; 藤本 望*

Annals of Nuclear Energy, 177, p.109314_1 - 109314_8, 2022/11

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

Power distribution plays a significant role in preventing the fuel temperature exceeds the safety limit of 1600$$^{circ}$$C in high-temperature gas-cooled reactors. The experiment to measure the power distribution in the graphite-moderated system was carried out with the Very High Temperature Reactor Critical Assembly facility. In the previous study, the power distribution in the VHTRC was calculated using a nuclear design code system based on diffusion calculation. The results showed a maximum discrepancy of up to 20% between the experiment and calculated values in the axial direction. The large discrepancy occurred near the boundary of fuel and reflector regions. This study describes the evaluation results of pin-wise power distribution of the VHTRC with the Monte Carlo MVP3 code. The calculation results were in good agreement with the measured results. In the axial direction, the discrepancy was less than 1% around the boundary of fuel and reflector regions.

論文

Calculation of shutdown gamma distribution in the high temperature engineering test reactor

Ho, H. Q.; 石井 俊晃; 長住 達; 小野 正人; 島崎 洋祐; 石塚 悦男; 後藤 実; Simanullang, I. L.*; 藤本 望*; 飯垣 和彦

Nuclear Engineering and Design, 396, p.111913_1 - 111913_9, 2022/09

 被引用回数:1 パーセンタイル:19.69(Nuclear Science & Technology)

Estimation of decay gamma distribution in a reactor core is essential for safely conducting various works after reactor shutdown such as periodic maintenance, shuffling fuel, removing spent fuel at the end of cycle, etc. Because of the dependency on the complex operating history of the reactor, attempting to calculate the decay gamma rays distribution in the core remains a challenge. This study showed a method to calculate the shutdown gamma distribution in the HTTR core by coupling a Monte-Carlo transport calculation code MCNP6 and an activation code ORIGEN2 to take advantage of spatial dependence and transportation abilities of MCNP6 and the detailed fission products tracking during burnup and cooling of ORIGEN2. As result, the three-dimensional shutdown gamma distribution in the HTTR core for different cooling times and spatial locations could be obtained accurately.

報告書

HTTRの核的パラメータの計算; 2021年度夏期休暇実習報告

五十川 浩希*; 直井 基将*; 山崎 誠司*; Ho, H. Q.; 片山 一成*; 松浦 秀明*; 藤本 望*; 石塚 悦男

JAEA-Technology 2022-015, 18 Pages, 2022/07

JAEA-Technology-2022-015.pdf:1.37MB

2021年度の夏期休暇実習において、HTTRの約10年の長期停止が臨界制御棒位置に与える影響及びMVPによるVHTRC-1炉心の遅発中性子割合の計算について検討した。この結果、長期停止が臨界制御棒位置に与える影響については、燃料内の$$^{241}$$Pu、$$^{241}$$Am、$$^{147}$$Pm、$$^{147}$$Sm、$$^{155}$$Gdの密度変化が影響して制御棒が4.0$$pm$$0.8cm引抜かれること、この計算値が測定値である3.9cmと近い値になることが明らかとなった。また、MVPによる遅発中性子割合の計算精度を確認するためVHTRC-1炉心について計算した結果、測定値を約10%過小評価することが明らかとなった。

論文

Reactor noise power-spectral analysis for a graphite-moderated and -reflected core, 3

左近 敦士*; 橋本 憲吾*; 佐野 忠史*; 中嶋 國弘*; 神田 峻*; 後藤 正樹*; 深谷 裕司; 沖田 将一朗; 藤本 望*; 高橋 佳之*

KURNS Progress Report 2021, P. 100, 2022/07

高温ガス炉の核特性を取得するための炉雑音解析技術の開発を京都大学臨界集合体(KUCA)を用い行っている。最新研究では、燃料集合体から55cm離れた検出器によりパワースペクトル密度の測定が行われた。しかしながら、即発中性子減衰定数は他の検出器から得られるものからの差異が発生した。そこで、本研究では炉外検出器によるパワースペクトル法による炉雑音解析を目的とする。

報告書

2020年度夏期休暇実習報告; HTTR炉心を用いた原子力電池に関する予備的検討; 核設計のための予備検討,3

石塚 悦男; 満井 渡*; 山本 雄大*; 中川 恭一*; Ho, H. Q.; 石井 俊晃; 濱本 真平; 長住 達; 高松 邦吉; Kenzhina, I.*; et al.

JAEA-Technology 2021-016, 16 Pages, 2021/09

JAEA-Technology-2021-016.pdf:1.8MB

2020年度の夏期休暇実習において、昨年度に引き続きHTTR炉心を原子力電池に見立てた場合の核的な予備検討として、MVP-BURNを用いて炉心の小型化について検討した。この結果、$$^{235}$$U濃縮度20%、54燃料ブロック(18$$times$$3層)炉心、半径1.6mのBeO反射体を使用すれば5MWで30年の連続運転が可能になることが明らかとなった。この小型炉心の燃料ブロック数は、HTTR炉心の36%に相当する。今後は、更なる小型化を目指して、燃料ブロックの材料を変更したケースについて検討する予定である。

報告書

MVP-BURNを用いた軸方向詳細モデルによるHTTRの燃焼特性解析

池田 礼治*; Ho, H. Q.; 長住 達; 石井 俊晃; 濱本 真平; 中野 優美*; 石塚 悦男; 藤本 望*

JAEA-Technology 2021-015, 32 Pages, 2021/09

JAEA-Technology-2021-015.pdf:2.74MB

MVP-BURNを用いてHTTR炉心の燃焼計算を行い、炉内温度分布を考慮した場合の影響とタリー領域分割を細分化した場合の影響を調べた。この結果、炉内温度分布を考慮した場合については、実効増倍率や主要核種密度に大きな影響がなかったこと、燃料ブロックごとの局所な$$^{235}$$U, $$^{239}$$Pu及び$$^{10}$$Bの物質量が最大で約6%、約8%及び約30%の差が生じたことが明らかとなった。また、タリー領域分割を細分化した場合については、実効増倍率への影響が0.6%$$Delta$$k/k以下と小さかったこと、黒鉛反射体の効果も含めた物質量の詳細分布、従来の計算より燃焼挙動を詳細に評価できることが明らかとなった。

論文

Nuclear data processing code FRENDY; A Verification with HTTR criticality benchmark experiments

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:37.09(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$$_{rm eff}$$ 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.

報告書

HTTR燃料セルモデルにおける可燃性毒物周辺のメッシュ効果

藤本 望*; 福田 航大*; 本多 友貴*; 栃尾 大輔; Ho, H. Q.; 長住 達; 石井 俊晃; 濱本 真平; 中野 優美*; 石塚 悦男

JAEA-Technology 2021-008, 23 Pages, 2021/06

JAEA-Technology-2021-008.pdf:2.62MB

SRACコードシステムを用いて可燃性毒物棒周辺のメッシュ分割がHTTR炉心の燃焼計算に与える影響を調べた。この結果、可燃性毒物棒内部のメッシュ分割は燃焼計算に大きな影響を与えないこと、実効増倍率は可燃性毒物棒周辺の黒鉛領域をメッシュ分割することで従来計算より測定値に近い値が得られることが明らかとなった。これにより、HTTR炉心の燃焼挙動をより適切に評価するには、可燃性毒物棒周辺黒鉛領域のメッシュ分割が重要になることが明らかとなった。

論文

Preparation for restarting the high temperature engineering test reactor; Development of utility tool for auto seeking critical control rod position

Ho, H. Q.; 藤本 望*; 濱本 真平; 長住 達; 後藤 実; 石塚 悦男

Nuclear Engineering and Design, 377, p.111161_1 - 111161_9, 2021/06

 被引用回数:3 パーセンタイル:37.09(Nuclear Science & Technology)

At high power operation of the HTTR, the control rod should be kept at the top of the active core for maintaining the optimized power distribution. It is important to calculate the control rod position each time the operating conditions change in order to ensure the safe operation of the reactor. Since the Monte Carlo code cannot change the core geometry such as control rod position during criticality and burnup calculation, the critical control rod position was determined by adjusting the control rods manually. Therefore, this study develops a new utility tool that seeks the control rod position automatically without any further handling procedures and waiting time. As a result, the determination of critical control rod position becomes simpler and the total time was also reduced significantly from about 5 days to less than 2 days. The calculated critical control rod position using the new tool also gives a good agreement with the experiment data.

論文

Effect of grain boundary on the friction coefficient of pure Fe under the oil lubrication

足立 望*; 松尾 泰貴*; 戸高 義一*; 藤本 幹也*; 日野 正裕*; 光原 昌寿*; 大場 洋次郎; 椎原 良典*; 梅野 宜崇*; 西田 稔*

Tribology International, 155, p.106781_1 - 106781_9, 2021/03

 被引用回数:11 パーセンタイル:65.79(Engineering, Mechanical)

Recent theoretical researches revealed that grain boundary affects tribological properties through the change in adsorption properties of additives in lubricants. Therefore, the effects of the grain boundary on the tribological properties were investigated by friction test of pure iron films under oil lubrication. We found that a friction coefficient decreases with increasing the fraction of the grain boundary in the lubricants forming chemisorbed film on sample surface. This suggests that the grain boundary enhances the formation of the chemisorbed films and reduces the friction coefficients by protecting the sample surface.

報告書

2019年度夏期休暇実習報告; HTTR炉心を用いた原子力電池に関する予備的検討; 核設計のための予備検討,2

石塚 悦男; 中島 弘貴*; 中川 直樹*; Ho, H. Q.; 石井 俊晃; 濱本 真平; 高松 邦吉; Kenzhina, I.*; Chikhray, Y.*; 松浦 秀明*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

2019年度の夏期休暇実習において、HTTR炉心を原子力電池に見立てた場合の核的な予備検討を実施し、MVP-BURNを用いて熱出力5MWで30年の連続運転が可能となる燃料の$$^{235}$$U濃縮度と可燃性毒物に関して検討した。この結果、$$^{235}$$U濃縮度が12%、可燃性毒物の半径及び天然ホウ素濃度が1.5cm及び2wt%の燃料が必要になることが明らかとなった。今後は、炉心の小型化について検討する予定である。

報告書

HTTR炉心解析における制御棒モデルの検討

長住 達; 松中 一朗*; 藤本 望*; 石井 俊晃; 石塚 悦男

JAEA-Technology 2020-003, 13 Pages, 2020/05

JAEA-Technology-2020-003.pdf:1.5MB

MVPコードを用いて、制御棒の幾何形状を実機の制御棒構造に近づけた詳細モデルを作成し、制御棒モデルの詳細化によるHTTRの核特性への影響について検討した。この結果、臨界制御棒位置は、制御棒モデルの詳細化により、従来のモデルと比べて11mm低くなり、実測値である1775mmに近づいた。また、制御棒先端のショックアブソーバーにより吸収された反応度は0.2%$$Delta$$k/kとなり、臨界制御棒位置にして14mmの差となることが分かった。さらに、制御棒の形状効果によるSRACコードの解析値に対する補正量は、制御棒モデルの詳細化とショックアブソーバーによる影響を考慮して、低温臨界時において反応度で-0.05%$$Delta$$k/k、臨界制御棒位置にして-3mmとなった。

論文

Promising neutron irradiation applications at the high temperature engineering test reactor

Ho, H. Q.; 本多 友貴*; 濱本 真平; 石井 俊晃; 高田 昌二; 藤本 望*; 石塚 悦男

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021902_1 - 021902_6, 2020/04

High temperature engineering test reactor (HTTR), a prismatic type of the HTGR, has been constructed to establish and upgrade the basic technologies for the HTGRs. Many irradiation regions are reserved in the HTTR to be served as a potential tool for an irradiation test reactor in order to promote innovative basic researches such as materials, fusion reactor technology, and radiation chemistry and so on. This study shows the overview of some possible irradiation applications at the HTTRs including neutron transmutation doping silicon (NTD-Si) and iodine-125 ($$^{125}$$I) productions. The HTTR has possibility to produce about 40 tons of doped Si-particles per year for fabrication of spherical silicon solar cell. Besides, the HTTR could also produce about 1.8$$times$$10$$^{5}$$ GBq/year of $$^{125}$$I isotope, comparing to 3.0$$times$$10$$^{3}$$ GBq of total $$^{125}$$I supplied in Japan in 2016.

論文

Conceptual design of direct $$^{rm 99m}$$Tc production facility at the high temperature engineering test reactor

Ho, H. Q.; 石田 大樹*; 濱本 真平; 石井 俊晃; 藤本 望*; 高木 直行*; 石塚 悦男

Nuclear Engineering and Design, 352, p.110174_1 - 110174_7, 2019/10

 被引用回数:2 パーセンタイル:19.31(Nuclear Science & Technology)

This study proposed a conceptual design of direct $$^{rm 99m}$$Tc production facility from a natural MoO$$_{3}$$ target at the high temperature engineering test reactor (HTTR). $$^{rm 99m}$$Tc is produced by a beta decay of $$^{99}$$Mo, which is formed via the $$^{98}$$Mo(n,$$gamma$$)$$^{99}$$Mo reaction. $$^{rm 99m}$$Tc is then extracted from the MoO$$_{3}$$ target by sublimation method to take advantage of the high temperature of the HTTR core. The foremost advantage of this concept is that the MoO$$_{3}$$ target is heated up inside the reactor without pulling out for external electric heating, and as a result, $$^{rm 99m}$$Tc could be extracted directly during irradiation. With 1 kg of MoO$$_{3}$$ target, the HTTR could produce about 6.8$$times$$10$$^{8}$$ MBq of $$^{rm 99m}$$Tc activity in comparison with 3.0$$times$$10$$^{8}$$ MBq of total $$^{rm 99m}$$Tc supplied in Japan in 2017.

報告書

HTTRの起動用中性子源の交換時期の推定

小野 正人; 小澤 太教; 藤本 望*

JAEA-Technology 2019-012, 15 Pages, 2019/09

JAEA-Technology-2019-012.pdf:2.83MB

HTTRでは、原子炉の起動及び広領域中性子検出器の計数率の確認を目的として、起動用中性子源$$^{252}$$Cfを用いているが、半減期が約2.6年と短いことから適切な時期に交換する必要がある。交換時期の推定には、広領域中性子検出器の「WRM計数率低」の警報発報を防ぐために、半減期のみならず、ゆらぎを考慮する必要がある。このため、広領域中性子検出器の計数率と標準偏差の関係式等から計数率の最小値を予測する手法を考案した。本手法を用いて広領域中性子検出器の計数率の変化を予測した結果、計数率が3.0cpsに低下するのが2022年、1.5cpsに低下するのが2024年となり、2024年までに交換を完了する必要があることが明らかとなった。

報告書

2018年度夏期休暇実習報告; HTTR炉心を用いた原子力電池に関する予備的検討; 核設計のための予備検討

石塚 悦男; 松中 一朗*; 石田 大樹*; Ho, H. Q.; 石井 俊晃; 濱本 真平; 高松 邦吉; Kenzhina, I.*; Chikhray, Y.*; 近藤 篤*; et al.

JAEA-Technology 2019-008, 12 Pages, 2019/07

JAEA-Technology-2019-008.pdf:2.37MB

2018年度の夏期休暇実習として、HTTR炉心を原子力電池に見立てた場合の核的な予備検討を実施した。この結果、熱出力2MWで約30年、3MWで約25年、4MWで約18年、5MWで約15年の運転が可能であるこが明らかとなった。また、熱的な予備検討として、自然循環冷却かつ可動機器のない発電システムを有する原子力電池のイメージを提案した。今後は、次年度の夏期休暇実習として更に検討を進め、原子力電池の成立性について検討する予定である。

論文

Feasibility study of large-scale production of iodine-125 at the high temperature engineering test reactor

Ho, H. Q.; 本多 友貴*; 濱本 真平; 石井 俊晃; 藤本 望*; 石塚 悦男

Applied Radiation and Isotopes, 140, p.209 - 214, 2018/10

 被引用回数:4 パーセンタイル:35.54(Chemistry, Inorganic & Nuclear)

The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors was investigated. A high-temperature engineering test reactor, which is located in Japan, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the $$^{125}$$I production could be maximized and the $$^{126}$$I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 180,000 GBq per year of $$^{125}$$I production.

論文

Feasibility study of new applications at the high-temperature gas-cooled reactor

Ho, H. Q.; 本多 友貴*; 濱本 真平; 石井 俊晃; 高田 昌二; 藤本 望*; 石塚 悦男

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

Besides the electricity generation and hydrogen production, HTGRs have many advantages for thermal neutron irradiation applications such as stable operation in longterm, large space available for irradiation target, and high thermal neutron economy. This study summarized the feasibility of new irradiation applications at the HTGRs including neutron transmutation doping silicon and I-125 productions. The HTTR located in Japan was used as a reference HTGR in this study. Calculation results show that HTTR could irradiate about 40 tons of doped Si particles per year for fabrication of spherical silicon solar cell. Besides, the HTTR could also produce about 1.8x105 GBq in a year of I-125, comparing to 3.0x103 GBq of total I-125 supplied in Japan in 2016.

論文

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

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).

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