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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:0.01(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:0 パーセンタイル:0.01(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.

論文

Consequence analysis of a postulated nuclear excursion in BWR spent fuel pool using 1/$$f^{beta}$$ spectrum model of randomization

Simanullang, I.; 山根 祐一; 菊地 丈夫; 外池 幸太郎

Annals of Nuclear Energy, 147, p.107675_1 - 107675_6, 2020/11

 被引用回数:2 パーセンタイル:24.76(Nuclear Science & Technology)

This study investigated a postulated criticality event that occurred in the fuel debris produced after the loss of the cooling function in a spent fuel pool. The 1/$$f^{beta}$$ spectrum model was introduced to overcome this situation considering the difficulties of predicting the random fuel debris distribution. The number of fissions in the first peak power was analyzed using the Nordheim Fuchs model. A total of 100 replicas were considered to investigate the fluctuation of the number of fissions. The results show that the number of fissions per volume varied from 3.78 $$times$$ 10$$^{19}$$ to 1.49 $$times$$ 10$$^{20}$$ (fissions/m$$^{3}$$). Moreover, the distribution of the number of fissions was positively skewed.

報告書

Calculation of decay heat by new ORIGEN libraries for high temperature engineering test reactor

Simanullang, I. L.*; 本多 友貴; 深谷 裕司; 後藤 実; 島崎 洋祐; 藤本 望*; 高田 昌二

JAEA-Technology 2015-032, 26 Pages, 2016/01

JAEA-Technology-2015-032.pdf:2.07MB

これまで高温工学試験研究炉の崩壊熱は、軽水炉のデータを基にしたShureの式やORIGEN計算で評価してきたが、厳密には軽水炉の中性子スペクトルと異なることから最適な評価方法を検討する必要がある。このため、黒鉛減速材量を変えた炉心の中性子スペクトルを用い、ORIGEN2コードで崩壊熱及び生成核種を計算して軽水炉の崩壊熱曲線と比較した。この結果、崩壊熱は、炉停止後1年程度であれば軽水炉と同様な値となったが、より長期になると差が顕著になり、$$^{90}$$Y, $$^{134}$$Cs, $$^{144}$$Pr, $$^{106}$$Rh, $$^{241}$$Am等が崩壊熱に大きく寄与することが明らかとなった。また、線量評価に関しては、冷却初期に$$^{241}$$Puが大きく影響することも明らかになった。

口頭

Effect of $$Gd_{2}O_{3}$$ to a fission number in the first pulse of a postulated nuclear excursion

Simanullang, I.; 山根 祐一; 外池 幸太郎

no journal, , 

沸騰水型軽水炉の使用済燃料プールで臨界事象が発生すると仮定して、可燃性毒物の初期装荷量が事象規模に与える効果を検討した。核分裂数を計算したところ、臨界安全設計で通常用いられる可燃性毒物を含まない二酸化ウラン燃料モデルは、保守的な結果を与えなかった。

口頭

Consequence analysis of postulated criticality in SFP using the randomized model of fuel debris

Simanullang, I.; 山根 祐一; 菊地 丈夫; 外池 幸太郎

no journal, , 

A postulated criticality event has been studied that occurred in fuel debris produced after the loss of cooling function in a spent fuel pool. The 1/f$$^{beta}$$ spectrum model was applied to deal with the random distribution composition of fuel debris. The results showed that the number of fissions per volume varying from 4.05 $$times$$ 10$$^{18}$$ to 1.6 $$times$$ 10$$^{19}$$.

口頭

Consequence analysis of a postulated nuclear excursion in BWR spent fuel pool using 1/$$f^{beta}$$ spectrum model of randomization

Simanullang, I.; 山根 祐一; 菊地 丈夫; 外池 幸太郎

no journal, , 

This study investigated a postulated criticality event that occurred in the fuel debris produced after the loss of the cooling function in a spent fuel pool. The 1/$$f^{beta}$$ spectrum model was introduced to overcome this situation considering the difficulties of predicting the random fuel debris distribution. The number of fissions in the first peak power was analyzed using the Nordheim Fuchs model. A total of 100 replicas were considered to investigate the fluctuation of the number of fissions. The results show that the number of fissions per volume varied from 3.78 $$times$$ 10$$^{19}$$ to 1.49 $$times$$ 10$$^{20}$$ (fissions/m$$^{3}$$). Moreover, the distribution of the number of fissions was positively skewed.

口頭

高温ガス炉用のORIGENライブラリ作成手法の検討

福原 克樹*; 藤本 望*; Simanullang, I. L.*; 深谷 裕司; Ho, H. Q.; 長住 達; 石井 俊晃; 濱本 真平; 石塚 悦男

no journal, , 

高温ガス炉燃料における核種生成消滅挙動の評価手法の確立を目指し、HTTRを対象として燃焼解析コードORIGENのライブラリ作成手法について検討している。本検討では、全炉心体系での中性子スペクトルの評価に基づき、燃料ブロック単位でのライブラリ作成を行った。そしてそのライブラリによるORIGENの燃焼結果とモンテカルロ法による全炉心燃焼結果を比較し、その妥当性を評価した。

口頭

Preparation method of ORIGEN2 library for high temperature gas-cooled reactors

Simanullang, I. L.*; 福原 克樹*; 守田 圭介; 深谷 裕司; Ho, H. Q.; 長住 達; 飯垣 和彦; 石塚 悦男; 藤本 望*

no journal, , 

The ORIGEN2 code has been used for fuel depletion calculations of many kinds of reactor fuels but there is no library for high temperature gas cooled reactors (HTGRs). A set of the ORIGEN2 library for the HTGR has been established to evaluate the burnup characteristics and compared its results with Monte Carlo burnup calculation results. The high temperature test engineering reactor (HTTR) is a block type fueled HTGRs in Japan. The ORIGEN2 libraries for the HTTR were prepared with neutron spectrums evaluated by pin-cell burnup calculations. To validate the calculation results of the ORIGEN2 code, a comparison between the burnup calculation using the Monte Carlo MVP-BURN code was conducted. One of the nuclide isotopes evaluated in that study was $$^{239}$$Pu. The amount of $$^{239}$$Pu calculated by ORIGEN2 with a library prepared by pin-cell burnup calculation was higher about 35% than that calculated by MVP-BURN. It can be seen that the ORIGEN2 library based on the pin cell model was not enough for evaluating the burnup characteristics of the HTTR. Therefore, an improvement was conducted by generating the ORIGEN2 library for HTTR based on the fuel block system and the whole system (core and reflector regions). The comparison between the ORIGEN2 results and MVP-BURN results was investigated. In the case of the ORIGEN2 library being generated based on the fuel block system, the maximum difference was about 6% compared to the mass of $$^{239}$$Pu calculated by MVP-BURN. Furthermore, the difference of $$^{239}$$Pu amount between the ORIGEN2 library and MVP-BURN became 2.4% when the library for ORIGEN2 was generated based on the whole system of HTTR.

口頭

Prediction of the operating control rod position of the HTTR with supervised machine learning

Ho, H. Q.; 長住 達; 島崎 洋祐; 濱本 真平; 飯垣 和彦; 後藤 実; Simanullang, I. L.*; 藤本 望*; 石塚 悦男

no journal, , 

During operation of the HTTR, hundreds of technical signals and operating conditions must be observed and evaluated to ensure safe operation of the reactor, for example reactor power, control rod position, coolant flow rate inlet/outlet, coolant temperature inlet/outlet, etc. The accumulated experiment data of the HTTR is not only important for the HTTR operation, but also for the basic development of the HTGR in general. Artificial intelligence (AI) and particularly machine learning (ML) are increasingly being used in various fields of research in modern science. They give the ability to make predictions as well as allow the extraction of key information about physical process from large datasets. Hence, there is a lot of potentials to apply AI and ML to predict the operating and safety parameters of the HTTR, and finally, a reactor simulator system for the HTTR could be expected by using the AI and ML algorithm. In this study, the control rod position of the HTTR is predicted based on ML without using the conventional neutronic codes. With the large accumulated data from operation history of the HTTR, the supervised ML with a linear regression algorithm was used. The linear regression algorithm finds a functional relationship between the input dataset (reactor power, burnup, etc.) and a reference dataset (control rod position), constructing a function that predicts control rod position from the other operation conditions. As result, the ML gives a good prediction of the HTTR control rod position with less than 5 difference compared to that in the experiment. This study is the initial step towards machine learning for research and analysis at the HTTR facility. With increasingly complicated experiments that create a large amount of data, ML is also expected to improve the design and safety analysis of the HTTR in the future.

口頭

High-temperature operation mode of HTTR for hydrogen production facility

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

no journal, , 

The conceptual design of a demonstration hydrogen production facility using heat supply from the high temperature engineering test reactor is being researched and developed at Japan Atomic Energy Agency. This facility produces hydrogen with a thermochemical water splitting Iodine-Sulphur process that requires high temperature heat to extract hydrogen efficiently. In order to achieve hydrogen with high efficiency, the helium outlet temperature should be as high as 950$$^{circ}$$C. Increasing the outlet temperature increases the reactor core temperature, and as a result the operation time decreases. If the operation time is reduced too much, it is not feasible to use the HTTR as a heat supply for the IS plant. Therefore, the purpose of this study is to estimate the operation time of the HTTR at high operation mode of 950$$^{circ}$$C to confirm whether it could supply long enough high temperature heat for the demonstration hydrogen IS plant. As result, the core temperature increase about by 50 to 100$$^{circ}$$C when the outlet temperature increases from 850 to 950$$^{circ}$$C. Although the increase of core temperature makes keff decrease by about 0.3 percent-delta-k per kelvin, the HTTR can still operate approximately 660EFPD. Therefore, it is possible to use the HTTR for long-term high-temperature heat supply to the demonstration hydrogen production IS plant.

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