※ 半角英数字
 年 ~ 
検索結果: 41 件中 1件目~20件目を表示


Initialising ...



Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...



Development of creep property equations of 316FR stainless steel and Mod.9Cr-1Mo steel for sodium-cooled fast reactor to achieve 60-year design life

鬼澤 高志; 橋立 竜太

Mechanical Engineering Journal (Internet), 6(1), p.18-00477_1 - 18-00477_15, 2019/02



Development of security and safety fuel for Pu-burner HTGR; Test and characterization for ZrC coating

植田 祥平; 相原 純; 後藤 実; 橘 幸男; 岡本 孝司*

Mechanical Engineering Journal (Internet), 5(5), p.18-00084_1 - 18-00084_9, 2018/10



ROSA/LSTF test on nitrogen gas behavior during reflux condensation in PWR and RELAP5 code analyses

竹田 武司; 大津 巌

Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08

We conducted an experiment focusing on nitrogen gas behavior during reflux condensation in PWR with ROSA/LSTF. The primary pressure was lower than 1 MPa under constant core power of 0.7% of volumetric-scaled (1/48) PWR nominal power. Steam generator (SG) secondary-side collapsed liquid level was maintained at certain liquid level above SG U-tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at certain constant amount. The primary pressure and degree of subcooling of SG U-tubes were largely dependent on amount of nitrogen gas accumulated in SG U-tubes. Nitrogen gas accumulated from outlet towards inlet of SG U-tubes. Non-uniform flow behavior was observed among SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in predictions of the primary pressure and degree of subcooling of SG U-tubes depending on number of nitrogen gas injection. We studied further non-uniform flow behavior through sensitivity analyses.


Development of a probabilistic risk assessment methodology against a combination hazard of strong wind and rainfall for sodium-cooled fast reactors

山野 秀将; 西野 裕之; 栗坂 健一

Mechanical Engineering Journal (Internet), 5(4), p.18-00093_1 - 18-00093_19, 2018/08



Development of probabilistic risk assessment methodology of decay heat removal function against combination hazard of low temperature and snow for sodium-cooled fast reactors

西野 裕之; 山野 秀将; 栗坂 健一

Mechanical Engineering Journal (Internet), 5(4), p.18-00079_1 - 18-00079_17, 2018/08

A probabilistic risk assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is aimed at developing a PRA methodology for the combination of low temperature and snow for a sodium-cooled fast reactor which uses the ambient air as its ultimate heat sink to remove decay heat under accident conditions. The annual exceedance probabilities of low temperature and of snow can be statistically estimated based on the meteorological records of temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., a clogged intake and/or outtake for a DHRS and for an emergency diesel generator, an unopenable door on necessary access routes due to accumulated snow, failure of intake filters due to accumulated snow, and possibility of water freezing in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around intakes and outtakes caused by loss of the access routes, this study has investigated effects of electric heaters installed around the intakes and outtakes as an additional countermeasure. By using the annual exceedance probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence caused by a snow and low temperature combination hazard is the failure of the electric heaters and the loss of the access routes for snow removal due to low temperature and snowfall which last for a day, and daily snowfall depth of 2 m/day.


Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor

今泉 悠也; 山田 文昭; 有川 晃弘*; 矢田 浩基; 深野 義隆

Mechanical Engineering Journal (Internet), 5(4), p.18-00083_1 - 18-00083_11, 2018/08



Evaluation of local damage to reinforced concrete panels subjected to oblique impact; Simulation analysis for evaluating perforation phenomena caused by oblique impact of deformable projectiles

西田 明美; 永井 穣*; 坪田 張二; Li, Y.

Mechanical Engineering Journal (Internet), 5(5), p.18-00087_1 - 18-00087_21, 2018/08



Validation of free-convective heat transfer analysis with JUPITER to evaluate air-cooling performance of fuel debris in dry method

上澤 伸一郎; 山下 晋; 柴田 光彦; 吉田 啓之

Mechanical Engineering Journal (Internet), 5(4), p.18-00115_1 - 18-00115_13, 2018/08

A dry method for fuel debris is proposed for decommissioning of TEPCO's Fukushima Daiichi NPS. However, air cooling performance has not yet been strictly evaluated for the fuel debris. We have evaluated the air-cooling performance of the fuel debris in the dry method by using JUPITER. Because the JUPITER can simulate melt relocation behavior of a reactor core, we can estimate air cooling performance for debris in consideration of the distribution and the structure of debris. In this paper, the validation of the free-convective heat transfer analysis of JUPITER were performed to evaluate the air-cooling performance of fuel debris in the dry method by using JUPITER. As the preliminary analysis, JUPITER was compared with OpenFOAM for simple configurations. The comparison proved that JUPITER can calculate the vertical temperature distribution as well as OpenFOAM on the condition of the lower heating amount. In the validation, JUPITER was compared with the heat transfer experiments of free convection in air adjacent to an upward-facing horizontal heating surface. The comparison proved that JUPITER was in good agreement with the experiment on the condition of the lower heating-surface temperature. The result indicated that JUPITER is a helpful numerical method to evaluate the free-convective heat transfer of the fuel debris in the dry method.


High-temperature oxidation of sheath materials using mineral-insulated cables for a simulated severe accident

中野 寛子; 広田 憲亮; 柴田 裕司; 武内 伴照; 土谷 邦彦

Mechanical Engineering Journal (Internet), 5(2), p.17-00594_1 - 17-00594_12, 2018/04

現在の軽水炉の計装システムは、原子炉運転と原子炉停止中の全ての状況を監視するために不可欠であるが、福島第一原子力発電所の重大事故のような状況では充分に機能しなかった。そのため、過酷事故時でも炉内の計測データを伝送可能な高温型MIケーブルを開発している。特に、過酷事故時の原子炉内は、窒素,酸素,水素,水蒸気のほかに核分裂生成物等が含まれた混合ガス雰囲気中に暴露されることから、シース材の早期破損が懸念される。本研究では、MIケーブル用シース材として選定したSUS316及びNCF600について、過酷事故を模擬した雰囲気(模擬大気,模擬大気/H$$_{2}$$O, I$$_{2}$$/CO/O$$_{2}$$/H$$_{2}$$O)中における高温酸化特性を調べた。その結果、模擬大気中または模擬大気/H$$_{2}$$O環境下におけるSUS316及びNCF600の両試料表面に均一な酸化皮膜が形成されるとともに、酸化速度を評価し、破断時間の予想が可能となった。一方、I$$_{2}$$が含まれている雰囲気では、試料表面の均一な酸化皮膜の形成だけでなく、局部腐食を引き起こす複雑な腐食挙動を示すことが分かった。


Study on a unified criterion for preventing plastic strain accumulation due to long distance travel of temperature distribution

岡島 智史; 若井 隆純; 川崎 信史

Mechanical Engineering Journal (Internet), 4(5), p.16-00641_1 - 16-00641_11, 2017/10

The prevention of excessive deformation by thermal ratcheting is important in the design of high-temperature components of fast breeder reactors (FBR). As a result of experimental study that simulated the fast breeder reactor vessel nearby the coolant surface, it was reported that long distance traveling of temperature distribution causes new type of thermal ratcheting, even if there is no primary stress. In this paper, we propose a simple screening criterion to prevent continuous accumulation of plastic strain derived from long distance traveling of temperature distribution. The major cause of this ratcheting mechanism is lack of residual stress that brings shakedown behavior at the center of yielding area. Because the residual stresses are derived from constraint against the elastic part, we focused on the distance from the center of yielding area to the elastic region. So, the proposed criterion restricts the axial length of the area with full-section yield state, which is the double of the above distance. We validated the proposed criterion based on finite element analyses using elastic-perfectly plastic material. As the result of the validation analyses, we confirmed that the accumulation of the plastic strain saturates before 2nd cycles in the cases that satisfy the proposed criterion, regardless of the shape of temperature distribution.


Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

小野田 雄一; 松場 賢一; 飛田 吉春; 鈴木 徹

Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06

For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow accident (ULOF) were preliminarily analyzed. The objective of this study is to demonstrate the integrity of the reactor vessel against the mechanical load induced by the energetics. Conservative energy production was assumed in order to confirm the robustness of the safety design of MONJU. Mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented. The mechanical energy, which were obtained by analyzing the expanding behavior of core materials after energetics, were about one order of magnitude below the thermodynamic work potential calculated by assuming isentropic expansion of the fuel vapor to one atmosphere, which was often used as an indicator to express the severity of the energetics. Structural integrity was then evaluated with coupled fluid-structure dynamics code using the obtained mechanical energy. No or very small circumferential residual strain of the reactor vessel was evaluated in most analytical cases, and even in the most conservative energy production case, the residual strain was only 0.008 % so that the integrity of the reactor vessel is maintained. The result obtained in the present study shows that MONJU has enough robustness against the mechanical load under energetics.


Design approach for mitigation of air ingress in high temperature gas-cooled reactor

佐藤 博之; 大橋 弘史; 中川 繁昭

Mechanical Engineering Journal (Internet), 4(3), p.16-00495_1 - 16-00495_11, 2017/06



Load and resistance factor design approach for seismic buckling of fast reactor vessels

高屋 茂; 佐々木 直人*; 浅山 泰; 神島 吉郎*

Mechanical Engineering Journal (Internet), 4(3), p.16-00558_1 - 16-00558_12, 2017/06



Sensitivity study on forest fire breakout and propagation conditions for forest fire hazard curve evaluations

岡野 靖; 山野 秀将

Mechanical Engineering Journal (Internet), 4(3), p.16-00517_1 - 16-00517_10, 2017/06



Development of numerical simulation method for melt relocation behavior in nuclear reactors; Validation and applicability for actual core structures

山下 晋; 徳島 二之*; 倉田 正輝; 吉田 啓之

Mechanical Engineering Journal (Internet), 4(3), p.16-00567_1 - 16-00567_13, 2017/06



Comparative study on neutronics characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

大釜 和也; Aliberti, G.*; Stauff, N. E.*; 大木 繁夫; Kim, T. K.*

Mechanical Engineering Journal (Internet), 4(3), p.16-00592_1 - 16-00592_9, 2017/06

Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using the Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, core characteristic parameters at the beginning of cycle were evaluated by the best estimate deterministic and stochastic methodologies of ANL and JAEA. The results obtained by both institutions show a good agreement with less than 200 pcm of discrepancy in the neutron multiplication factor, and less than 3% of discrepancy in the sodium void reactivity, Doppler reactivity, and control rod worth. The results by the stochastic and deterministic approaches were compared in each party to investigate impacts of the deterministic approximation and to understand potential variations in the results due to different calculation methodologies employed. Impacts of the nuclear data libraries were also investigated using a sensitivity analysis methodology.


Evaluation of sodium pool fire and thermal consequence in two-cell configuration

高田 孝; 大野 修司; 田嶋 雄次*

Mechanical Engineering Journal (Internet), 4(3), p.16-00577_1 - 16-00577_11, 2017/06



Influence of inlet velocity condition on unsteady flow characteristics in piping with a short elbow under a high-Reynolds-number condition

小野 綾子; 田中 正暁; 小林 順; 上出 英樹

Mechanical Engineering Journal (Internet), 4(1), p.16-00217_1 - 16-00217_15, 2017/02



Effect of dissolved gases on mechanical property of AISI 304 and 316 stainless steels under high temperature and pressure water

武内 伴照; 中野 寛子; 上原 聡明; 土谷 邦彦

Mechanical Engineering Journal (Internet), p.95 - 96, 2016/09



Benchmark analyses of probabilistic fracture mechanics for cast stainless steel pipe

北条 公伸*; 林 翔太郎*; 西 亘*; 釜谷 昌幸*; 勝山 仁哉; 眞崎 浩一*; 永井 政貴*; 岡本 年樹*; 高田 泰和*; 吉村 忍*

Mechanical Engineering Journal (Internet), 3(4), p.16-00083_1 - 16-00083_16, 2016/08


41 件中 1件目~20件目を表示