Development of Transient and Safety Analysis Method for ATF With SiC

Based on lessons learned from the Fukushima Daiichi nuclear power plant accident, pursuit of accident tolerant fuel (ATF) has been discussed by many institutions in the world. Toshiba identified a silicon carbide (SiC) ceramic as the most promising material for accident tolerant fuel. Since SiC has less active characteristics in the presence of high temperature water steam (H2O) and is expected to be tolerant of severe accident conditions. Moreover, SiC has a smaller neutron absorption cross-section which is advantageous feature in terms of neutron economy. Zirconium alloys (Zry) are one of the main structural materials in LWR core. In high temperature H2O environment under severe accident conditions, Zry rapidly reacts with H2O and oxidation reaction accompanied by release of hydrogen gas occurs. Since SiC may inhibit the progress of oxidation reaction compared to Zry metal alloys, hydrogen and heat generation is expected to decrease in the case of core uncovered accident conditions. In order to confirm the advantage of SiC over Zry as core materials, transient analysis and safety analysis are carried out. For transient analysis, analyses of temperature behavior of cladding at plant transient condition are carried out with best-estimate transient analysis code. This analysis confirmed the effect of physical properties differences between SiC and Zry on cladding temperature behavior. Moreover to indicate the effectiveness of SiC under the core uncovered condition with oxidation reaction, safety analysis with latest “MAAP” code is carried out and the whole plant behavior during severe accident sequence is simulated. This analysis showed the effectiveness of SiC to mitigate the oxidation reaction. As the result of these analyses, the advantage of SiC over Zry can be perceived. And also, future challenges of SiC application as ATF can be clarified through these analyses.

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Identification of Failure Modes Under Design Extension Conditions

Volume 1B: Codes and Standards ◽

10.1115/pvp2015-45381 ◽

2015 ◽

Cited By ~ 1

Author(s):

Naoto Kasahara ◽

Izumi Nakamura ◽

Hideo Machida ◽

Hitoshi Nakamura ◽

Koji Okamoto

Keyword(s):

High Temperature ◽

Nuclear Power ◽

Failure Modes ◽

Low Cycle Fatigue ◽

External Pressure ◽

Lessons Learned ◽

Local Failure ◽

Severe Accident ◽

Mitigation Measures ◽

Loading Mode

As the important lessons learned from the Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure became essential against severe accident and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premise under design extension conditions such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, best estimation for these failure modes are required for preparing countermeasures and management. Therefore, this study focused on identification of failure modes under design extension conditions. To observe ultimate failure behaviors of structures under extreme loadings, new experimental techniques were adopted with simulation materials such as lead and lead-antimony alloy, which has very small yield stress. Postulated failure modes of main components under design extension conditions were investigated according three categories of loading modes. The first loading mode is high temperature and internal pressure. Under this mode, ductile fracture and local failure were investigated. At the structural discontinuities, local failure may become dominant. The second is high temperature and external pressure loading mode. Buckling and fracture were investigated. Buckling occurs however hardly break without additional loads or constraints. The last loading is excessive earthquake. Ratchet deformation, collapse, and fatigue were investigated. Among them, low-cycle fatigue is dominant.

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Boron chemistry during transportation in the high temperature region of a boiling water reactor under severe accident conditions

Journal of Nuclear Science and Technology ◽

10.1080/00223131.2019.1671913 ◽

2019 ◽

Vol 57 (3) ◽

pp. 291-300 ◽

Cited By ~ 1

Author(s):

Shuhei Miwa ◽

Gaku Takase ◽

Junpei Imoto ◽

Shunichiro Nishioka ◽

Naoya Miyahara ◽

...

Keyword(s):

High Temperature ◽

Temperature Region ◽

Boiling Water ◽

Severe Accident ◽

Boiling Water Reactor ◽

High Temperature Region ◽

Water Reactor ◽

Boron Chemistry ◽

Accident Conditions

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Prediction of Hydrogen Distribution Under Severe Accident Conditions With CFD and Lumped Parameter Codes

Volume 3: Thermal-Hydraulics; Turbines, Generators, and Auxiliaries ◽

10.1115/icone20-power2012-54499 ◽

2012 ◽

Cited By ~ 2

Author(s):

Tobias Szabó ◽

Stefan Benz ◽

Frank Kretzschmar ◽

Peter Royl ◽

Thomas Jordan

Keyword(s):

Source Term ◽

Severe Accident ◽

Test Bed ◽

Hydrogen Distribution ◽

Water Steam ◽

Lumped Parameter ◽

Accident Conditions ◽

Energy Balances ◽

Mass And Energy Balances ◽

Good Agreement

In case of a severe accident, the containment is the ultimate barrier to the environment. Therefore, reliable simulations tools for containment thermal hydraulics, including hydrogen distribution are indispensable. We simulated the behavior of the containment atmosphere under severe accident conditions with a postulated source term of water, steam and hydrogen. We used a detailed 3D CFD code (GASFLOW) and a lumped parameter code (MELCOR) in order to compare and assess their modeling capabilities. A simplified generic containment including all important components was used as a test bed. We analyzed the calculated pressure histories, mass and energy balances, convective flow as well as steam and hydrogen distributions. Integral values were modeled in good agreement by both codes. The overall flow was reasonably predicted. However we observed discrepancies in the calculated steam and hydrogen concentrations.

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RESULTS OF A PHENOMENA IDENTIFICATION AND RANKING TABLE (PIRT) EXERCISE FOR A SEVERE ACCIDENT IN A SMALL MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR

CNL Nuclear Review ◽

10.12943/cnr.2019.00006 ◽

2019 ◽

Vol 8 (2) ◽

pp. 159-169

Author(s):

David William Hummel ◽

Yu-Shan Chin ◽

Andrew Prudil ◽

Anthony Williams ◽

Eugene Masala ◽

...

Keyword(s):

High Temperature ◽

Current Knowledge ◽

Safety Analysis ◽

Severe Accident ◽

Specific Interest ◽

Worst Case ◽

Severe Accidents ◽

Small Modular Reactor ◽

Physical Phenomena ◽

High Temperature Gas

Canada has attracted specific interest from developers of nonwater-cooled small modular reactor (SMR) technologies, including concepts based on high-temperature gas-cooled reactors (HTGRs). It is anticipated that some research and development (R&D) will be necessary to support safety analysis and licensing of these reactors in Canada. The Phenomena Identification and Ranking Table (PIRT) process is a formalized method in which a panel of experts identifies which physical phenomena are most relevant to the reactor safety analysis and how well understood these phenomena are. The PIRT process is thus a tool to assess current knowledge levels and (or) predictive capabilities of models, thus providing direction to a focused R&D program. This paper summarizes the results of a PIRT process performed by a panel of experts at Canadian Nuclear Laboratories for a limiting or “worst-case” accident scenario at a generic HTGR-type SMR. Suggestions are given regarding the highest priority R&D items to support severe accidents analysis of these reactors.

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Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

Journal of Nuclear Materials ◽

10.1016/j.jnucmat.2017.02.021 ◽

2017 ◽

Vol 487 ◽

pp. 229-237 ◽

Cited By ~ 21

Author(s):

Y. Yano ◽

T. Tanno ◽

H. Oka ◽

S. Ohtsuka ◽

T. Inoue ◽

...

Keyword(s):

High Temperature ◽

Tensile Properties ◽

Severe Accident ◽

Ods Steel ◽

Accident Conditions ◽

High Temperature Tensile ◽

High Temperature Tensile Properties ◽

Ultra High Temperature

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Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes

Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory ◽

10.1115/icone26-81923 ◽

2018 ◽

Author(s):

Hideki Horie ◽

Yutaka Takeuchi ◽

Kenya Takiwaki ◽

Fumie Sebe ◽

Kazuo Kakiuchi ◽

...

Keyword(s):

High Temperature ◽

Oxidation Reaction ◽

Nuclear Power ◽

Reaction Rate ◽

Hydrogen Generation ◽

Reactor Core ◽

The Other ◽

Severe Accident ◽

Fuel Cladding ◽

The Difference

Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry. Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction. In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case. On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.

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