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Journal Articles

Development of a mixed oxide fuel pin performance analysis code "CEDAR"; Models and analyses of fuel pin irradiation behavior

Uwaba, Tomoyuki; Mizuno, Tomoyasu; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*

Nuclear Engineering and Design, 280, p.27 - 36, 2014/12

 Times Cited Count:7 Percentile:37.75(Nuclear Science & Technology)

A deterministic computer code CEDAR has been developed to analyze irradiation behaviors of a mixed-oxide fuel pellet pin in a FBR. The FEM was incorporated into the mechanical calculation part of the code for properly analyzing stress-strain status in the fuel pellet and cladding, and mechanical interaction between the fuel pellet and cladding. The code features mechanistic analyses of irradiation behaviors of a fuel pin by integrating a lot of models to analyze major irradiation phenomena, thus expressing actual fuel pin irradiation behaviors. Analysis capabilities of the code were validated by calculations of fuel pellet temperatures, fractional fission gas releases of fuel pins and fuel pin cladding diametral strain profiles. The mechanisms of the fuel pin irradiation behaviors such as redistribution of Americium, PCMI and JOG formation were interpreted from the code analyses for the actual irradiation test fuel pins.

JAEA Reports

Irradiation behavior analyses of oxide fuel pins for SFR high breeding cores

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-012, 13 Pages, 2013/06

JAEA-Technology-2013-012.pdf:2.34MB

A mixed oxide fuel pin concept with annular pellets and an ODS cladding is a possible driver fuel for commercialized Sodium-cooled fast reactor (SFR) core. This fuel concept was considered with low breeding ratio as a standard, break-even breeding cores and cores with high breeding ratio (high breeding cores). Some calculations of fuel pin irradiation performance of (U,Pu) oxide fuel and minor actinides bearing oxide fuel were conducted by a fuel performance analysis code CEDAR developed in JAEA to understand the steady state irradiation behavior of fuel pins for the cores with high breeding ratio. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at end of life (EOL) were evaluated. Those results show that the MOX fuel pin having the specifications and irradiation conditions used in this investigation would be irradiated moderately up to approximately 250 GWd/t with well integrity.

JAEA Reports

Fast reactor fuel pin behavior analyses in a LOF type transient event

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-011, 10 Pages, 2013/06

JAEA-Technology-2013-011.pdf:2.02MB

In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, the fast reactor fuel pin performance code CEDAR was used for calculation. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross & Stoute type gap conductance model and constant gap conductance model used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross & Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of the former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross & Stoute type gap conductance model which is thought to be realistic.

JAEA Reports

Fuel temperature analyses of metallic fuel pins for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-010, 17 Pages, 2013/06

JAEA-Technology-2013-010.pdf:2.46MB

Metallic fuel, U-Pu(TRU)-Zr is a fuel candidate for Sodium-cooled fast reactor (SFR) selected as a possible promising future nuclear reactor system in Generation-IV international forum (GIF). Design studies were performed in the Japanese feasibility study on commercialized fast reactor cycle system, and the irradiation behavior of metallic fuel is under investigation through analytical fuel performance code calculations with preliminary analytical models. Some calculations of U-Pu(TRU)-Zr fuel irradiation performance were conducted by a simplified calculation grogram developed in JAEA. Axial profile of fuel pin centerline temperature calculated by using effective fuel thermal conductivity where sodium ingress into fuel was considered fits well with actual fuel micro structures after the irradiation. The effective fuel thermal conductivity with sodium ingress is suitable for the irradiation behavior investigation.

JAEA Reports

Fuel temperature analyses at overpower of metallic fuel pin for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-009, 12 Pages, 2013/06

JAEA-Technology-2013-009.pdf:1.3MB

Metallic fuel, U-Pu(TRU)-Zr is a fuel candidate for Sodium-cooled fast reactor (SFR) selected as a possible promising future nuclear reactor system in Generation-IV international forum (GIF). Design studies were performed in the Japanese feasibility study on commercialized fast reactor cycle system, and the irradiation behavior of metallic fuel is under investigation through analytical fuel performance code calculations with preliminary analytical models. As fuel temperature analyses at overpower events are also major interest, some calculations of U-Pu(TRU)-Zr fuel irradiation performance were conducted by a simplified calculation program developed in JAEA. The calculated fuel temperature at the maximum power of overpower events, 110-120% of steady state power, was around 1100K in maxim. It is clear that this temperature was low enough to avoid fuel melting in the event.

JAEA Reports

Irradiation behavior analyses of oxide fuel pins for startup core of a demonstration SFR

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-007, 17 Pages, 2013/05

JAEA-Technology-2013-007.pdf:1.69MB

As a swelling resistant austenitic steel, PNC316, is candidate cladding tube material of the first core of demonstration Sodium-cooled fast reactor (SFR). The irradiation behavior of an annular MOX fuel pin with (U,Pu) oxide fuel contained in PNC 316 cladding was evaluated by a fuel performance analysis code CEDAR developed in JAEA. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin keeps its integrity at least up to 100 GWd/t of peak burnup.

JAEA Reports

Irradiation behavior analyses of MA bearing oxide fuel pin for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-006, 17 Pages, 2013/05

JAEA-Technology-2013-006.pdf:1.72MB

As a fuel concept for commercialized Sodium-cooled fast reactor (SFR) system, minor actinides (MA) bearing oxide fuel with oxide dispersion strengthened martensitic steel (ODS) cladding was considered under homogeneous TRU recycling strategy. The MA content is calculated to be around 5% of heavy metal in case of trans-uranium (TRU) feed from light water reactor (LWR) spent fuel during the transition phase from LWR to fast reactor era. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at end of life (EOL) were evaluated by fuel performance analytical code CEDAR developed in JAEA to investigate the irradiation behavior of annular MOX fuel pins with (U,Pu) oxide fuel and Am bearing oxide fuel under a high burnup condition. Also, the profiles of pressure on the cladding inner surface and the cladding deformation after irradiation were evaluated.

JAEA Reports

Irradiation behavior analyses of oxide fuel pin for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-005, 17 Pages, 2013/05

JAEA-Technology-2013-005.pdf:1.7MB

A mixed oxide fuel pin concept with annular pellets and an oxide dispersion strengthened martensitic steel (ODS) cladding is a possible driver fuel for commercialized Sodium-cooled fast reactor (SFR) core. The capability of annular MOX fuel pins with (U,Pu) oxide fuel and Am bearing oxide fuel under a high burnup condition was evaluated by a fuel performance analysis code CEDAR developed in JAEA. The fuel temperature profiles, fuel and cladding deformation profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin had enough safety margin to fuel melting under the irradiation. Also, the profiles of pressure on the cladding inner surface and the cladding deformation after irradiation were evaluated. Those results show that the gap of the fuel pin at fabrication had enough width not to occur the considerable fuel-cladding mechanical interaction (FCMI).

JAEA Reports

Irradiation behavior analyses of metallic fuel pins for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

JAEA-Technology 2013-004, 16 Pages, 2013/05

JAEA-Technology-2013-004.pdf:1.52MB

In order to be an alternative concept to the conventional concept consisting of mixed oxide (MOX) fuel, Metallic fuel, U-Pu(TRU)-Zr metallic fuel slug and ODS cladding were considered for Sodium-cooled fast reactor (SFR) cycle system. The capability of the U-Pu(TRU)-Zr metallic fuel with ODS cladding under a high burnup condition was calculated and conducted by a simplified calculation grogram developed in JAEA. The fuel temperature profiles, gap width profiles, and radial temperature distribution at EOL were evaluated. Those results show that the fuel pin had enough safety margin to fuel melting under the irradiation. Evaluation of the profiles of plenum gas pressure and the cladding deformation after irradiation shows that the fuel pin had enough plenum volume not to cause considerable cladding deformations by plenum gas pressure. In case of 0.4% Am bearing fuel, calculation result shows that fuel centerline temperature becomes high, but increase from U-Pu-Zr fuel is insignificant.

Journal Articles

Study on the mechanism of diametral cladding strain and mixed-oxide fuel element breaching in slow-ramp extended overpower transients

Uwaba, Tomoyuki; Maeda, Seiichiro; Mizuno, Tomoyasu; Teague, M. C.*

Journal of Nuclear Materials, 429(1-3), p.149 - 158, 2012/10

 Times Cited Count:2 Percentile:77.38(Materials Science, Multidisciplinary)

Cladding strain caused by fuel/mechanical interaction (FCMI) was evaluated for mixed-oxide fuel elements subjected to 70-90% slow-ramp extended overpower transient tests in EBR-II. Calculated transient-induced cladding strains were correlated with cumulative damage fractions (CDFs) using cladding strength correlations. In a breached high-smeared density solid fuel element with low strength cladding, cladding thermal creep strain was significantly increased to approximately half the transient-induced cladding strain due to the tertiary creep when the CDF was close to the breach criterion (= 1.0). In low-smeared density annular fuel elements, FCMI load was significantly mitigated and resulted in little cladding strain. The CDFs of the annular fuel elements were lower than 0.01 at the end of the overpower transient, indicating a substantial margin to breach. A substantial margin to breach was also maintained in a high-smeared density fuel element with high strength cladding.

Journal Articles

Correlations among FBR core characteristics for various fuel compositions

Maruyama, Shuhei; Oki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu

Journal of Nuclear Science and Technology, 49(6), p.640 - 654, 2012/06

 Times Cited Count:2 Percentile:77.38(Nuclear Science & Technology)

This study shows the good correlations in FBR core characteristics, and find out the mechanism of their correlations with the aid of sensitivity analyses. It has been clarified that Doppler coefficient turns to have the correlations with the other core characteristics by considering the constraint of the criticality requirement for fuel composition variations. The finding of the correlations makes easy to specify the ranges of core reactivity control and core safety properties which are important for core design in determining core specification and performance. It gives significant information for FBR core design in the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. By using this index and the correlations, the core characteristic variations can be estimated for various fuel compositions without repeating core calculations.

Journal Articles

U-Pu-Zr metallic fuel core and fuel concept for SFR with a 550$$^{circ}$$C core outlet temperature

Naganuma, Masayuki; Ogata, Takanari*; Mizuno, Tomoyasu

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

In the FaCT project, a design study of the metallic fuel SFR has been executed as secondary candidate. The primary interest is to achieve a core outlet temperature of 550 $$^{circ}$$C. However, the metallic fuel has a drawback that the maximum temperature of the cladding inner surface is limited to 650 $$^{circ}$$C to avoid liquid phase formation. To overcome this problem, JAEA has developed and studied the advanced core concept with single Pu-enrichment and 2 radial regions of heavy metal density. In this paper, the core and fuel design study for the middle-scale SFR applying this core concept are discussed. In addition, for the practical application of the metallic fuel to the SFR with high outlet temperature, it is necessary to expand irradiation experience under the high cladding temperature condition. Therefore, JAEA and CRIEPI planned an irradiation test of the metallic fuel in Joyo as a collaborative program. In this paper, the outline and current status of the irradiation test are reported.

Journal Articles

Fast reactor core design considerations from proliferation resistance aspects

Kawashima, Katsuyuki; Ogawa, Takashi; Oki, Shigeo; Okubo, Tsutomu; Mizuno, Tomoyasu

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

Sodium-cooled fast reactor core design considerations are made to improve the proliferation resistance by focusing on the plutonium generated in the UO$$_{2}$$ blanket in the frame of the Fast Reactor Cycle Technology Development (FaCT) project. The appropriate design and treatments of the UO$$_{2}$$ blanket help to reduce the intrinsic proliferation potentials. Based on the 1500 MWe FaCT reference core, the three different cores (radial blanket-free core, the core with the low-enriched MOX fuel, and the core with MA-doped UO$$_{2}$$ fuel) are configured to meet the provisional proliferation resistance criteria as well as the core performance targets.

Journal Articles

Power distribution skewing effects on fuel temperature during TOP in a large commercial-base fast reactor

Kawashima, Katsuyuki; Okubo, Tsutomu; Mizuno, Tomoyasu

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12

As part of the FaCT Project, the study is focusing on to evaluate the effects of power distribution skewing on the fuel temperature at the TOP state in the large commercial-base Fast Reactor (JSFR), in order to help streamline the safety evaluation methods of the JSFR. It is shown that the considerations of the power distribution skewing increase the average fuel temperature rise, leading to more negative reactivity feedback which decreases the rate of hot pin temperature rise during TOP, and the peak linear power is significantly mitigated by local feedback reactivities due to temperature rise in the fuel assemblies adjacent to the withdrawn control rod. A conventional calculation model which neither considers the power distribution skewing effect nor the local feedback reactivities effect gives the conservative results of the peak linear power in the TOP event.

Journal Articles

A Preliminary assessment of the adoption of innovative technologies in the Fast Reactor Cycle Technology Development (FaCT) project in Japan

Sato, Koji; Kotake, Shoji; Fujita, Yuji; Mizuno, Tomoyasu

Energy Procedia, 7, p.140 - 145, 2011/09

 Times Cited Count:2 Percentile:12.38

JAEA has been implementing the FaCT project in cooperation with electric utilities toward the commercialization of fast reactor cycle system before 2050. In this FaCT project, many innovative technologies with technical challenges are actively used in order to provide significant improvements in economic competitiveness, enhancement of safety and reliability, sustainability, and nonproliferation. The work of deciding on the adoption of innovative technologies by the end of JFY2010 is in progress. This paper describes current preliminary assessment results.

Journal Articles

Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

Uwaba, Tomoyuki; Ito, Masahiro*; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, B. J.*; Wootan, D. W.*; Carmack, J.*

Journal of Nuclear Materials, 412(3), p.294 - 300, 2011/05

 Times Cited Count:8 Percentile:38.3(Materials Science, Multidisciplinary)

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence of about 39$$times$$10$$^{26}$$n/m$$^{2}$$ as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

Journal Articles

Evaluation of creep damage and diametral strain of fast reactor MOX fuel pins irradiated to high burnups

Uwaba, Tomoyuki; Sogame, Motomu; Ito, Masahiro*; Mizuno, Tomoyasu; Donomae, Takako; Katsuyama, Kozo

Journal of Nuclear Science and Technology, 47(8), p.712 - 720, 2010/08

 Times Cited Count:5 Percentile:59.02(Nuclear Science & Technology)

In determining lifetime criteria of fast reactor fuel pins, creep damage due to fission gas pressure on mixed-oxide fuel pins with austenitic stainless steel cladding successfully irradiated to high burnups (120 GWd/t or higher pin averaged burnup) was evaluated. The degree of creep damage of these fuel pins was expressed as cumulative damage fractions (CDFs), defined so that cladding breaching occurs when the CDF exceeds 1.0. The obtained CDFs for typical high temperature fuel pins were on the order of 10$$^{-4}$$-10$$^{-2}$$ at the end of irradiation, indicating that these fuel pins had large safety margins against breaching due to creep damage. In order to investigate the factors that govern the lifetime of fuel pins, pin diametral increase as well as CDF were predicted in cases of extended burnups from 120 GWd/t onward, and then were compared with tentatively determined limit values. The predicted pin diametral increase reached its limit value earlier than the CDF because of a significant increase in the cladding void swelling, suggesting that lifetimes of fuel pins with austenitic stainless steel cladding are practically governed by the diametral increase rather than by the creep damage.

Journal Articles

The Impact of americium target in-core loading on reactivity characteristics and ULOF response of sodium-cooled MOX FBR

Yamaji, Akifumi; Kawashima, Katsuyuki; Oki, Shigeo; Mizuno, Tomoyasu; Okubo, Tsutomu

Nuclear Technology, 171(2), p.153 - 160, 2010/08

 Times Cited Count:4 Percentile:64.88(Nuclear Science & Technology)

The homogeneous MA loading core with 3wt% MAs is used as a reference design to evaluate the impact of the americium target in-core loading (20wt% MAs) on reactivity characteristics and ULOF response of sodium-cooled MOX-FBR. The Am target loading method of this study can flatten core radial reactivity worth distributions and effectively reduce reactivity insertion into the core during ULOF. As the result, the core power increase speed during ULOF is reduced. The maximum fuel temperature of the target region does not become particularly high compared with that of the inner core and it is much lower than the melting point. It is promising from the viewpoints of the reactivity characteristics and ULOF response.

Journal Articles

Minor actinide-bearing oxide fuel core design study for the JSFR

Naganuma, Masayuki; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Kotake, Shoji*

Nuclear Technology, 170(1), p.170 - 180, 2010/04

 Times Cited Count:9 Percentile:39.6(Nuclear Science & Technology)

In FaCT project, sodium-cooled fast reactor with mixed-oxide fuel was selected as the primary candidate. Present study focused on effects of TRU composition on the design of JSFR. In a transitional stage from LWR to FBR, there is possibility for the JSFR fuel to have high MA content due to recycle of LWR spent fuel. That affects core and fuel designs (core reactivity, material property and so on). Thus, to evaluate the effects quantitatively, design studies for the JSFR cores with two TRU compositions were conducted, one was FBR multi-recycle composition with about 1wt% of MA and the other was LWR recycle one, for which 3wt% of MA was assumed as a typical composition. The results showed that the change from FBR multi-recycle composition to LWR recycle one leads to 10% increase of sodium void reactivity, 1-2% decrease of linear power limit and 5% extension of gas plenum length. As a result, effects of TRU composition on the core and fuel designs were indicated to be benign.

JAEA Reports

Fuel and core design studies on metal fuel sodium-cooled fast reactor, 3; Joint research report for JFY2007&2008

Okano, Yasushi; Kobayashi, Noboru*; Ogawa, Takashi; Oki, Shigeo; Naganuma, Masayuki; Okubo, Tsutomu; Mizuno, Tomoyasu; Ogata, Takanari*; Ueda, Nobuyuki*; Nishimura, Satoshi*

JAEA-Research 2009-025, 105 Pages, 2009/10

JAEA-Research-2009-025.pdf:10.45MB

A metal fuel core has specific features on high heavy metal density, hard neutron spectrum, and efficient neutron utilization. Enlarged applicable design envelops would improve core performances and features: higher breeding ratio, compacted reactor core, and, smaller amount of Pu-fissile inventory. A joint study on "Reactor Core and Fuel Design of Metal Fuel Core of Sodium Cooled Fast Reactor" by Japan Atomic Energy Agency and Central Research Institute of Electric Power Industry has been conducted during Japanese fiscal years of 2007 and 2008. This report shows the results on (1) the study on applicable design ranges of metal fuel specifications, (2) the study on conceptual core designs for high breeding ratio, and (3) the safety study on metal fuel core designed in the Fast Reactor Cycle Technology Development (FaCT) Project.

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