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

Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka

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

https://doi.org/10.1016/j.nucengdes.2023.112443

Characteristics of allowable axial cracks for pressurized pipes governed by limit load criteria

Hasegawa, Kunio; Li, Y.; Udyawar, A.*; Lacroix, V.*

International Journal of Pressure Vessels and Piping, 204, p.104952_1 - 104952_7, 2023/08

https://doi.org/10.1016/j.ijpvp.2023.104952

When axial cracks were detected in pipes, failure stresses for high toughness pipes are estimated using the Limit Load Criteria. The allowable stresses for the cracked pipes are derived from the combination of the failure stresses and safety factors. The allowable sizes of crack depths and lengths are determined from the allowable stresses. From the comparison of the allowable and failure stresses for through-wall cracks, the allowable cracks are not uniform. They can be separated into three different characteristics, i) leak-before-break (LBB) and crack growth stability, ii) non-LBB and crack growth stability and iii) non-LBB and crack growth instability. Inspectors and users should pay special attention to allowable cracks with the third characteristic to prevent unexpected failure, particularly for thin-wall pipes. The allowable crack depths and lengths that require special attention can be expressed by appropriate equations.

Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurised water reactor primary coolant under irradiation, 2; Evaluation of electrochemical corrosion potential

Hata, Kuniki; Hanawa, Satoshi; Chimi, Yasuhiro; Uchida, Shunsuke; Lister, D. H.*

Journal of Nuclear Science and Technology, 60(8), p.867 - 880, 2023/08

https://doi.org/10.1080/00223131.2022.2127955

One of the major subjects for evaluating the corrosive conditions in the PWR primary coolant was to determine the optimal hydrogen concentration for mitigating PWSCC without any adverse effects on major structural materials. As suitable procedures for evaluating the corrosive conditions in PWR primary coolant, a couple of procedures, i.e., water radiolysis and ECP analyses, were proposed. The previous article showed the radiolysis calculation in the PWR primary coolant, which was followed by an ECP study here. The ECP analysis, a couple of a mixed potential model and an oxide layer growth model, was developed originally for BWR conditions, which was extended to PWR conditions with adding Li (Na) and H effects on the anodic polarization curves. As a result of comparison of the calculated results with INCA in-pile-loop experiment data as well as other experimental data, it was confirmed that the ECPs calculated with the coupled analyses agreed with the measured within 100mV discrepancies.

Analytical study for low ground contact ratio of buildings due to the basemat uplift using a three-dimensional finite element model

Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.; Ota, Akira*; Sonobe, Hideaki*; Ino, Susumu*; Ugata, Takeshi*

Mechanical Engineering Journal (Internet), 10(4), p.23-00026_1 - 23-00026_11, 2023/08

https://doi.org/10.1299/mej.23-00026

In the seismic evaluation of nuclear facility buildings, basemat uplift-the phenomenon during which the bottom of the basemat of a building partially rises from the ground owing to overturning moments during earthquakes-is a very important aspect because it affects not only structural strength and integrity, but also the response of equipment installed in the building. However, there are not enough analytical studies on the behavior of buildings with a low ground contact ratio due to basemat uplift during earthquakes. In this study, we conducted a simulation using a three-dimensional finite element model from past experiments on basemat uplift; further, we confirmed the validity of this approach. In order to confirm the difference in the analytical results depending on the analysis code, the simulation was performed under the same analytical conditions using the three analysis codes, which are E-FrontISTR, FINAS/STAR and TDAPIII, and the obtained analysis results were compared. Accordingly, we investigated the influence of the difference in adhesion on the structural response at low ground contact ratio. In addition, we confirmed the effects of significant analysis parameters on the structural response via sensitivity analysis. In this paper, we report the analytical results and insights obtained from these investigations.

Data report of ROSA/LSTF experiment IB-HL-01; 17% hot leg intermediate break LOCA with totally-failed high pressure injection system

Takeda, Takeshi

JAEA-Data/Code 2023-007, 72 Pages, 2023/07

JAEA-Data-Code-2023-007.pdf:3.24MB

An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.

Revision of allowable planar flaw tables of ASME B&PV Code Section XI for ferritic steel materials

Dulieu, P.*; Lacroix, V.*; Hasegawa, Kunio

Proceedings of ASME 2023 Pressure Vessels & Piping Conference (PVP 2023) (Internet), 7 Pages, 2023/07

When defects were found during in-service inspection in nuclear components, the ASME Code Section XI provides allowable flaw sizes to assess the flaw severity. For ferritic steel materials, the sizes of allowable planar flaws given in Table IWB-3510-1 were determined by the stress intensity factors. The objective of this methodology is including some basic criteria to prevent plastic collapse and brittle failure. As far as the prevention from plastic collapse, a uniform limit load reduction is considered whatever the flaw aspect ratios. For the prevention of brittle failure, a reference surface flaw configuration is defined to derive a reference stress intensity factor. This methodology is applied to surface flaws with various aspect ratios. It is also coherently applied to subsurface flaws considering the proximity of the flaw to the surface as an additional parameter. Finally, a revision of the allowable planar flaw Table IWB-3510-1 of ASME Code Section XI is proposed.

Effect of calcium on niobium solubility in alkaline solutions

Ohira, Saki; Abe, Takeyasu; Iida, Yoshihisa

Radiochimica Acta, 111(7), p.525 - 531, 2023/07

https://doi.org/10.1515/ract-2022-0115

The solubility of Nb in calcium alkaline solutions is one of the important parameters in safety assessment of intermediate-depth disposal which are assumed to use cementitious materials. Nb solubility and solubility-limiting solid phases of Nb in these systems remain unclear. The oversaturation solubility experiments were performed systematically in the 0.001-0.1 M CaCl solutions under alkali conditions, and the characterization of precipitated solid phase controlling Nb solubility was conducted. The negative dependence of Nb solubilities on pH and Ca concentration was observed in solubility experiments, the molar ratio of Nb to Ca of precipitated solid phase was 0.66. The pH and Ca dependence of Nb solubilities was reproduced by the reaction with Nb aqueous species Nb(OH) and Ca-Nb oxide with the molar ratio of Nb to Ca 0.66, e.g., CaNbO(am).

Study on borehole sealing corresponding to hydrogeological structures by groundwater flow analysis

Sawaguchi, Takuma; Takai, Shizuka; Sasagawa, Tsuyoshi; Uchikoshi, Emiko*; Shima, Yosuke*; Takeda, Seiji

MRS Advances (Internet), 8(6), p.243 - 249, 2023/06

https://doi.org/10.1557/s43580-023-00528-y

In the intermediate depth disposal of relatively high-level radioactive waste, a method to confirm whether the borehole for monitoring is properly sealed should be developed in advance. In this study, groundwater flow analyses were performed for the hydrogeological structures with backfilled boreholes, assuming sedimentary rock area, to understand what backfill design conditions could prevent significant water pathways in the borehole, and to identify the confirmation points of borehole sealing. The results indicated the conditions to prevent water pathways in the borehole and BDZ (Borehole Disturbed Zone), such as designing the permeability of bentonite material less than or equal to that of the host rock, and grouting BDZ.

Proposition of confirmation items on the borehole sealing for the disposal of radioactive waste

Murakami, Hiroaki; Nishiyama, Nariaki; Takeuchi, Ryuji; Iwatsuki, Teruki

Oyo Chishitsu, 64(2), p.60 - 69, 2023/06

https://doi.org/10.5110/jjseg.64.60

In order to confirm the quality control items for borehole closure in radioactive waste disposal projects, in-situ borehole sealing tests using bentonite material were conducted. As a result, the closure performance was successfully demonstrated by comparing the data of water injection tests conducted before and after the installation of the closure material. However, the breakthrough was observed after closing, probably due to high differential pressure applied to the seal section. Thus, it is important to ascertain throughout the entire operation that the borehole is adequately closed. The placement and specifications of the closure material should be determined according to the hydrogeological structure in the borehole. The confirmation items to use bentonite for sealing material are identified to be: to consider swelling and density loss in the borehole; to place the planned depth using appropriate emplacement technique; to be placed without damage to seals when use some backfilling materials, considering effect of permeability on adjacent seals.

Accident sequence precursor analysis of an incident in a Japanese nuclear power plant based on dynamic probabilistic risk assessment

Kubo, Kotaro

Science and Technology of Nuclear Installations, 2023, p.7402217_1 - 7402217_12, 2023/06

https://doi.org/10.1155/2023/7402217