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報告書

燃料挙動解析コードパッケージFEMAXIの機能拡充; 軽水炉燃料の反応度事故時挙動解析モジュールRANNSの開発と検証

田崎 雄大; 宇田川 豊

JAEA-Data/Code 2024-012, 76 Pages, 2024/12

JAEA-Data-Code-2024-012.pdf:9.25MB

日本原子力研究開発機構(以下、「原子力機構」という。)では、軽水炉燃料の通常運転時及び過渡条件下の挙動評価を目的として、燃料挙動解析コードFEMAXIを開発してきた。2019年3月には、同コードとして初めて体系的な検証と性能評価を行ったFEMAXI-8を公開し、以降も様々な改良を続けている。並行して、原子力機構では2000年代より、設計基準事故(DBA)解析用のブランチとしてRANNSモジュールの開発を進めてきた。RANNSは、DBA条件、ここでは主に反応度事故(RIA)の様な非常に急峻な過渡に対しても燃料挙動を追跡できるよう、特に計算の安定性を重視しつつ、このような過渡挙動を適切に予測する上で重要な沸騰熱伝達、粒界分離を伴うFPガス放出、破壊力学指標に基づく被覆管破損判定などを特有のモデルとして備えている。本報告では、これら事故時挙動解析向けモデルの解説やプログラムの設計・構造におけるFEMAXIとの関係に加え、原子力機構が研究炉NSRRを用いて実験を実施し、蓄積してきた膨大なRIA実験データによる大規模検証の結果を示し、同モジュールの総合的なRIA解析性能を評価している。RANNSモジュールの公開に当たっては、パッケージ化されたFEMAXI/RANNSとしてユーザへ提供する予定であり、これにより広い条件での燃料挙動を解析することが可能となる。また、検証解析を通じて一定の性能が確認されたモデルパラメータセットも本報告内で提示しており、これを参照することで、これまで公開してきたFEMAXI-8とユーザビリティは殆ど変わることなく、また解析者の力量に大きく依存することなく、事故時挙動解析の実行が可能である。

論文

Bayesian statistical model for cladding high-temperature burst under loss-of-coolant accident conditions

田崎 雄大; 成川 隆文; 宇田川 豊

Journal of Nuclear Science and Technology, 61(10), p.1349 - 1359, 2024/10

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

This study developed a probabilistic determination model with respect to cladding high-temperature burst conditions based on the Bayesian statistical method to reasonably evaluate fuel behaviors under loss-of-coolant accident conditions, including fuel fragmentation, relocation, and dispersal. The candidate models were based on the widely accepted empirical model established based on nonirradiated fuel cladding data. Explanatory variables were added to improve the applicability of these models with respect to irradiated materials and generalization performance. The posterior predictive distribution of each candidate model was evaluated using Bayesian estimation comprising 238 sets of high-temperature burst test data. The generalization performance was evaluated using information criteria. The results of model evaluation showed improved predictive performance by considering the effect of hydrogen content. A comparison with burnup as an alternative explanatory variable confirmed that hydrogen content was the better parameter and other burnup-associated effects, such as irradiation hardening of the metal matrix and oxide growth (reduction of the metal matrix), were less dominant under burst conditions.

論文

Fission gas dynamics test; Development of reactivity-initiated-accident testing technique devoted to investigation of fission gas release kinetics

三原 武; 浦野 建太; 宇田川 豊

Proceedings of TopFuel 2024 (Internet), 9 Pages, 2024/10

To promote a better understanding of the fission gas behavior during a reactivity-initiated accident (RIA) and its role in the thermo-mechanical loading in the fuel cladding, the Fission Gas Dynamics (FGD) test program has been developed in the framework of the JAEA and Institute for Radiological Protection and Nuclear Safety (IRSN) cooperation. The concept of the FGD test is to understand the effect of fission gas release during a RIA test through transient measurement of the pressure inside a rigid chamber, which contains the test fuel, with its minimum deformation against pressure increase. Since the internal pressure sensor of strain gauge (SG) type used in previous RIA-simulated Nuclear Safety Research Reactor (NSRR) tests is strongly affected by gamma and/or neutron field in the NSRR core, we adopted a new pressure sensor using a linear variable differential transducer (LVDT) for accurate pressure measurement with higher stability against pulse irradiation. JAEA has conducted the first NSRR-FGD test (FGD-1) on high-burnup fuel with doped pellets. In advance, difference in pressure response between LVDT-type and SG-type sensors was carefully examined as performance measure of the LVDT-type sensor. The response delay of the LVDT-type sensor compared with the SG-type one was estimated to be about 1.5 ms when the pressure increasing rate exceeded above 20 MPa/s. In the FGD-1 test, the LVDT-type pressure sensor detected a pressure rise of about 100 MPa/s just after the pulse irradiation, which confirmed the capability of this FGD testing technique to study the kinetics of rapid fission gas release during the simulated RIA conditions.

論文

A Study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

Li, F.; 三原 武; 宇田川 豊

Journal of Nuclear Science and Technology, 61(9), p.1265 - 1275, 2024/09

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

In a part of the reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the fracture pattern change conditions. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of 0.5-1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio is greater than 0.75 and a concurrent hoop strain $$<$$ 1.7% at the failure instant.

論文

The Effect of a cyclic bending load on the bending resistance of ballooned, ruptured, and oxidized Zircaloy-4 cladding

Li, F.; 成川 隆文; 宇田川 豊

Journal of Nuclear Science and Technology, 61(8), p.1036 - 1047, 2024/08

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

The seismic resistance of fuel cladding during the long-term core cooling after loss-of-coolant accidents (LOCAs) was investigated by performing cyclic four-point bending tests (4PBTs) of up to 1000 cycles with fresh fuel cladding samples that experienced integral thermal shock test, simulating LOCA conditions, including ballooning, rupture, oxidation, and quench. 4PBTs were performed on the samples that survived the quenching process. The results showed that up to 1000 cycles and 5.8 Nm of cyclic loading moment, there was no apparent effect on the bending fracture limit of the fuel cladding under the 4PBT. The scatter of the bending fracture limit for a given equivalent cladding reacted (ECR) evaluated by the Baker-Just oxidation rate equation (BJ-ECR) is attributed to two primary factors: first, the difference between the prescribed and the actual oxidation behavior, confirmed by comparing the BJ-ECR and the ECR evaluated based on metallographic observation (M-ECR), and second, the variated shape of the rupture-opening area after the integral thermal shock test. The strength of the alpha phase-dominant zone near the rupture opening seems to contribute to the bending fracture limit.

論文

Uncertainty analysis of model selection based on information criterion; A Case study of a probability estimation model for fuel cladding tube fracture during LOCA

成川 隆文; 宇田川 豊

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03

Information criteria such as a widely applicable information criterion (WAIC) and a widely applicable Bayesian information criterion (WBIC) enable the selection of models with high predictive accuracy and data fit, yet these criteria come with inherent uncertainties as they are statistical measures. To evaluate the uncertainty in model selection based on these information criteria, we performed numerical experiments using the bootstrap method, which is a resampling technique, on models for estimating the fracture probability of fuel cladding tubes during loss-of-coolant accidents (LOCAs). By calculating WAIC and WBIC for each of 10,000 bootstrap samples, we evaluated the dependency of model selection on these samples. Our key findings reveal that: (1) Sample-derived variation in information criteria was significantly greater than variability between models, underscoring the importance of assessing uncertainty from samples. (2) The Log-probit model, developed in our previous study, was selected as the optimal model for its superior predictive performance and data fit, despite the inherent uncertainties associated with WAIC and WBIC. (3) The presence of outliers at the fracture/non-fracture boundary of fuel cladding tubes may negatively impact the information criteria, suggesting the need for careful consideration when including such data in model parameter estimation.

論文

Modeling of the P2M past fuel melting experiments with the FEMAXI-8 code

Mohamad, A. B.; 宇田川 豊

Nuclear Technology, 210(2), p.245 - 260, 2024/02

 被引用回数:2 パーセンタイル:49.11(Nuclear Science & Technology)

In the Power to Melt and Maneuverability (P2M) project, a simulation exercise on two past power ramp experiments xM3 on medium burn-up rod and HBC4 on high burn-up rod were performed with the fuel performance code FEMAXI-8 to investigate the fuel behavior under high power and high-temperature conditions toward centerline fuel melting. In order to treat fuel melting, empirical melting temperature models have been incorporated into the FEMAXI-8 code. The present analysis gave reasonable predictions not only on cladding deformation but also on the fuel melting behavior of the HBC4 rod, in which the UO$$_{2}$$ liquidus temperature was reached during the transient. On the other hand, model improvement appeared to be needed for a more accurate treatment of fuel melting behavior of the xM3 rod, in which fuel center temperature reached solidus line, whereas may not reached liquidus line. A reasonable agreement of estimated FGR with the measurement suggested that the high temperature FGR at the given conditions are essentially temperature dependent phenomenon: rate-limited primarily by thermally activated elementary processes such as fission gas diffusion.

論文

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 三原 武; 垣内 一雄; 宇田川 豊

Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01

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

A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5$$^{TM}$$ cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.

論文

Oxidation and embrittlement behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

成川 隆文; 近藤 啓悦; 藤村 由希; 垣内 一雄; 宇田川 豊; 根本 義之

Journal of Nuclear Materials, 587, p.154736_1 - 154736_8, 2023/12

 被引用回数:2 パーセンタイル:49.11(Materials Science, Multidisciplinary)

To evaluate the oxidation and embrittlement behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions, we conducted isothermal oxidation and ring-compression tests on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens. Further, we discussed the loss of coolable geometry of the reactor core loaded with the FeCrAl-ODS cladding tubes under LOCA conditions, using data from the ring-compression tests in this study and the integral thermal shock tests from our previous study. The results reveal that oxidation kinetics of the FeCrAl-ODS cladding tube at 1523 K is four orders of magnitude lower than that of a conventional Zircaloy cladding tube, which highlights the exceptional oxidation resistance of the FeCrAl-ODS cladding tube. The breakaway oxidation of the FeCrAl-ODS cladding tube was observed at 1623 K for durations equal to or exceeding 6 h, and melting was observed at 1723 K. The ring-compression and the integral thermal shock tests indicate that, depending on the oxidation time, the ductile to brittle transition threshold - as determined by the ring-compression test - exists between 1623 K and 1723 K. Meanwhile, the fracture threshold - established through the integral thermal shock test - falls between 1573 K and 1673 K. Therefore, taking a conservative approach based on available data, the fracture and non-fracture results from the integral thermal shock tests can define the lower and upper boundaries of the threshold for the loss of coolable geometry of the reactor core during a LOCA.

論文

Hierarchical Bayesian modeling to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

成川 隆文; 濱口 修輔*; 高田 孝*; 宇田川 豊

Nuclear Engineering and Design, 411, p.112443_1 - 112443_12, 2023/09

 被引用回数:1 パーセンタイル:27.70(Nuclear Science & Technology)

For realizing a highly reliable fracture limit evaluation of fuel cladding tubes during loss-of-coolant accidents (LOCAs) in light-water reactors, we developed a method to quantify the fracture limit uncertainty of high-burnup advanced fuel cladding tubes. This method employs a hierarchical Bayesian model that can quantify uncertainty even with limited experimental data. The fracture limit uncertainty was quantified as a probability using the amount of oxidation (Equivalent cladding reacted: ECR) and the initial hydrogen concentration (the hydrogen concentration in the fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. We divided the regression coefficients of this model into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences among various types of fuel cladding tubes. This hierarchical structure enabled us to quantify the fracture limit uncertainty through the effective use of prior knowledge and data, even for high-burnup advanced fuel cladding tubes with a small number of data points. The fracture limits representing a 5% fracture probability with 95% confidence of the high-burnup advanced fuel cladding tubes evaluated by the hierarchical Bayesian model were higher than 15% ECR for the initial hydrogen concentrations of up to 700-900 wtppm and restraint loads below 535 N. These fracture limits were comparable to the limit of the unirradiated Zircaloy-4 cladding tube, indicating that the burnup extension and use of the advanced fuel cladding tubes do not significantly lower the fracture limit of fuel cladding tubes. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data, instead of the binary data, depending on the condition of the fuel cladding tube specimens after performing the LOCA-simulated test, thereby increasing the amount of information in the data.

論文

Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

成川 隆文; 近藤 啓悦; 藤村 由希; 垣内 一雄; 宇田川 豊; 根本 義之

Journal of Nuclear Materials, 582, p.154467_1 - 154467_12, 2023/08

 被引用回数:3 パーセンタイル:65.16(Materials Science, Multidisciplinary)

To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200-300 K higher than that of the Zircaloy cladding tube, and the FeCrAl-ODS cladding tube's maximum circumferential strain was smaller than or equal to the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of $$sim$$5000 N, which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at 1473 K for 1 h. The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at 1673 K and fractured after abnormal oxidation at 1573 K for 1 h. Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below 1473 K, where no melting or abnormal oxidation occurs.

論文

Effects of azimuthal temperature distribution and rod internal gas energy on ballooning deformation and rupture opening formation of a 17 $$times$$ 17 type PWR fuel cladding tube under LOCA-simulated burst conditions

古本 健一郎; 宇田川 豊

Journal of Nuclear Science and Technology, 60(5), p.500 - 511, 2023/05

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

In order to contribute to better modeling and evaluation of fuel fragmentation, relocation, and dispersal expected under loss of coolant accident (LOCA) conditions, LOCA-simulated cladding burst experiments were performed on as-received nonirradiated 17 $$times$$ 17 type Zircaloy-4 cladding specimens that were internally pressurized. The experiments were designed to terminate at burst occurrence to focus on ballooning and rupture opening formation and to investigate the effects of various factors. The postburst cladding hoop strain decreased with the increase in azimuthal temperature distribution (ATD) of the cladding, as found previously. The rupture opening size increased with the increase in ATD and the increase in energy of the pressurized gas stored inside the pressure boundary of the test sample system. Comparison with the existing database, which included tests on irradiated rods containing fuel pellets, suggested that formation of the rupture opening was influenced by the characteristic behavior of high burnup fuels, such as limited gas migration in the cladding tube due to fuel-cladding bonding and interaction of the ejected fuel fragments with the cladding tube.

論文

Behavior of high-burnup BWR UO$$_{2}$$ fuel with additives under reactivity-initiated accident conditions

三原 武; 垣内 一雄; 谷口 良徳; 宇田川 豊

Journal of Nuclear Science and Technology, 60(5), p.512 - 525, 2023/05

 被引用回数:1 パーセンタイル:15.53(Nuclear Science & Technology)

Fuels with additives are expected to provide enhanced fuel performance in fission gas retention owing to their large grain size, which elongates fission gas migration path. To investigate behavior of the fuels during a reactivity-initiated accident (RIA), RIA-simulated experiments OS-1 and LS-4 were performed on ADOPT (chromia- and alumina-doped UO$$_{2}$$) fuel of 64 GWd/t and chromia-doped UO$$_{2}$$ fuel of 48 GWd/t, respectively. The OS-1 rod failed at a fuel enthalpy increase of 160 J/g due to pellet-cladding mechanical interaction failure, which was the lowest failure limit among the test results ever obtained at the NSRR on high-burnup fuels from 40 to 65 GWd/tU. Comparison of the hydride morphologies in the cladding metallic layer between the rods subjected to the past NSRR tests suggests the contribution of radially oriented hydrides during base irradiation to the low failure limit. The LS-4 rod survived for a peak fuel enthalpy increase of 549 J/g, which resulted in cladding deformation of $$sim$$2.4% in the residual hoop strain and FGR of 1.4%-6.1%. Whereas the low fission gas release exhibits the effect of additives, the cladding deformation is within the range explained by the deformation mechanism essentially identical to those recognized for high-burnup undoped fuels.

論文

Engineering formulation of the irradiation growth behavior of zirconium-based alloys for light water reactors

垣内 一雄; 天谷 政樹; 宇田川 豊

Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01

 被引用回数:2 パーセンタイル:15.53(Materials Science, Multidisciplinary)

The irradiation growth behavior of coupon specimens prepared from improved Zr-based alloys for light-water reactor fuel cladding, which have various additive elements and fabrication conditions, was investigated by conducting an irradiation test at 573 and 593 K under typical PWR coolant conditions up to a fast-neutron fluence of $$approx$$7.8$$times$$10$$^{21}$$ (n/cm $$^{2}$$, E $$>$$1 MeV) in the Halden reactor in Norway. Based on the dimensional change data measured at interim and final inspections, the amounts of irradiation growth of the improved Zr-based alloys were formulated from the viewpoint of engineering. The trends of the parameters which express the effects of additive elements on irradiation growth behavior were in good agreement with those previously reported, and it was found that the amount of irradiation growth can be expressed by using a summation rule of the effect of each additive element on irradiation growth.

論文

Evaluation of anisotropic elastic and plastic parameters of Zircaloy-4 fuel cladding from biaxial stress test data and their application to a fracture mechanics analysis

Li, F.; 三原 武; 宇田川 豊

Journal of Nuclear Science and Technology, 59(12), p.1455 - 1464, 2022/12

 被引用回数:2 パーセンタイル:30.61(Nuclear Science & Technology)

The mechanical properties of fuel cladding near the elastic limit are essential in considering its failure limit during a pellet-cladding mechanical interaction phase under reactivity-initiated accident (RIA) conditions. The mechanical properties of a Zircaloy-4 cladding tube, such as orthotropic elasticity and anisotropic constants for Hill's plasticity law, were evaluated based on the biaxial stress test data, focusing on the equivalent plastic strain up to ~2.5%. Samples with various fabrication conditions, such as cold-worked, recrystallized, and stress relieved after cold-work with Q-factors of 2, 3, and 4 were investigated. The cold worked samples and recrystallized samples showed high yield stress and Young's modulus, respectively. The evaluated mechanical properties of the stress relieved samples revealed a limited impact of Q-factors on mechanical behavior, including their anisotropic feature. The derived mechanical properties were applied to evaluate the fracture mechanics parameter, J-integral, based on failure limit data from biaxial-expansion-due-to-compression tests on precracked tubes. This evaluation produced systematically lower J-integral values of the stress relieved tube than previously evaluated based on the failure limit data from in-pile RIA-simulated tests.

論文

Hierarchical Bayes model to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under LOCA conditions

成川 隆文; 濱口 修輔*; 高田 孝*; 宇田川 豊

Proceedings of Asian Symposium on Risk Assessment and Management 2022 (ASRAM 2022) (Internet), 11 Pages, 2022/12

To realize a more reliable safety evaluation of loss-of-coolant accidents (LOCAs) in light-water-reactors, we developed a quantification method of the fracture limit uncertainty of high-burnup advanced fuel cladding tubes using a hierarchical Bayes model that can quantify uncertainty even when experimental data are limited. The fracture limit uncertainty was quantified as a probability using the amount of oxidation and the initial hydrogen concentration (the hydrogen concentration in fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. The hierarchical Bayes model was developed by dividing the regression coefficients into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences between types of fuel cladding tubes. Using the developed model, we showed that the fracture limits of the high-burnup advanced fuel cladding tubes tended to be on average equal to or higher than that of an unirradiated conventional fuel cladding tube. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data depending on the condition of the fuel cladding tube specimens after the LOCA-simulated test instead of the binary data, thereby increasing the amount of information in each data.

論文

Mechanical property evaluation with nanoindentation method on Zircaloy-4 cladding tube after LOCA-simulated experiment

垣内 一雄; 山内 紹裕*; 天谷 政樹; 宇田川 豊; 北野 剛司*

Proceedings of TopFuel 2022 (Internet), p.409 - 418, 2022/10

In order to examine the influence of cladding microstructural changes upon the mechanical property of the fuel cladding under LOCA conditions in a more direct and quantitative manner, the nanoindentation method has been applied to Zircaloy-4 cladding specimens after LOCA simulated tests (about 1473 K, ECR 20%, quench at 973 K after slow cooling); results for two specimens taken from the rupture opening part and secondary hydriding part were compared. In addition to hardness and Young's modulus, the plastic work fraction that corresponds to the relative ductility was evaluated from the load-displacement curve. The plastic work fraction at the secondary hydriding part was found to be obviously lower than that at the rupture opening part and closer to that in $$alpha$$-Zr(O) layers beneath the outer surface. This result from the nanoindentation method agrees with the conventional knowledge about low ductility at the secondary hydriding part.

報告書

ナノインデンテーション法によるLOCA模擬試験後ジルカロイ被覆管の機械特性評価(共同研究)

垣内 一雄; 宇田川 豊; 山内 紹裕*

JAEA-Research 2022-001, 21 Pages, 2022/06

JAEA-Research-2022-001.pdf:1.84MB

冷却材喪失事故(LOCA)時想定される被覆管脆化の主たる要因は、高温酸化に伴う金属層中酸素濃度の増大とこれに起因する微細組織の変化である。被覆管が破裂した場合には、燃料棒内に侵入した水蒸気によって生じる被覆管内面の酸化及びこれに伴う燃料棒内水素分圧の上昇の結果、破裂開口部からやや離れた軸方向位置で局所的な水素吸収が起こり(二次水素化)、二次水素化部では水素脆化による延性低下も重畳する。これら微細組織の変化がLOCA条件下における燃料棒の機械特性に及ぼす影響をより詳細かつ定量的に把握するため、LOCA模擬試験後試料の破裂開口部及び二次水素化部の延性評価にナノインデンテーション法を適用した。硬さやヤング率に加えて、押込み荷重-変位曲線から算出される塑性仕事割合を評価したところ、二次水素化部の金属層(prior-$$beta$$相)における塑性仕事割合は、被覆管外周のZrO$$_{2}$$層と$$alpha$$-Zr(O)層に近い水準であり、破裂開口部に比べて酸素濃度が低いにもかかわらず、水素の影響により有意に延性が低下していることが示唆された。

論文

Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

垣内 一雄; 天谷 政樹; 宇田川 豊

Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06

 被引用回数:7 パーセンタイル:70.05(Nuclear Science & Technology)

In order to understand the dimensional stability of the fuel rod during long-term use in commercial LWRs, an irradiation growth testing in the Halden reactor of Norway was conducted on various fuel cladding materials including the improved Zr alloy. In this paper, the effect of hydrogen, which was absorbed in the cladding tube due to corrosion, on the irradiation growth behavior was evaluated. Comparison between the specimens with or without pre-charged hydrogen revealed that the effect of hydrogen absorption, accelerating irradiation growth, became significant when the hydrogen content exceeded the hydrogen solubility limit in the corresponding irradiation temperature. Analysis based on this understanding derived growth acceleration effect (0.06$$pm$$0.01)%/100 ppm, whose denominator is defined as the amount of absorbed hydrogen involved in hydride precipitation under irradiation as a relevant parameter.

論文

Development of fission gas release model for MOX fuel pellets with treatment of heterogeneous microstructure

田崎 雄大; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03

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

This study develops a new fission gas release (FGR) model for mixed oxide (MOX) fuels with a fundamentally heterogeneous microstructure. The model adopted in FEMAXI-8 was applied to irradiation Instrumented Fuel Assembly (IFA)-626 and 702 tests in which two types of MOX fuels had different heterogeneity in their microstructure, while the other spec were similar. Upon analyzing these fuels, the original FGR model predicted lower FGR from the fuel with a remarkably heterogeneous microstructure than the other MOX fuel. This estimation contradicts the experimental observation. However, the new FGR model improved the consistency because of the early release of fission gas from Pu agglomerate region, and showed issues for aiming further improvement. Therefore, the above results confirmed a certain validity of the developed model for studying heterogeneity effect.

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