Japanese Evaluated Nuclear Data Library Version 3 Revision-2: JENDL-3.2

Nakagawa, Tsuneo; Shibata, Kozo; Chiba, Satoshi; Fukahori, Tokio; Nakajima, Yoshikazu; Kikuchi, Yasuyuki; Kawano, Toshihiko; Kanda, Yoshiteru; Ohsawa, Takaaki; Matsunobu, Hiroyuki; Kawai, Masayoshi; Zukeran, Atsushi; Watanabe, Takashi; Igarasi, S.; Kosako, Kazuaki; Asami, Tetsuo

doi:10.1080/18811248.1995.9731849

Cited by 427 publications

(44 citation statements)

References 24 publications

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“…It should be noted that the neutronics aspect is the main consideration in the present study. The neutronics cell calculation [7] was performed by using PIJ (collision probability method code) routine of SRAC 2002 code [8], with nuclear data library is JENDL-3.2 [9]. The SRAC code with JENDL-3.2 library consists of 107 energy groups, where 48 thermal groups and 74 fast groups with 15 overlapping groups.…”

Section: Methodsmentioning

confidence: 99%

Comparative Studies on Plutonium and Minor Actinides Utilization in Small Molten Salt Reactors with Various Powers and Core Sizes

et al. 2015

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Molten salt reactor (MSR) has many advantages such as safety improvement and ability for waste burning. In the present study, the use of plutonium and minor actinides (MA) in small MSRs with various powers and core sizes has been investigated. For the small MSRs with 50 MWth, 100 MWth, 150 MWth, 200 MWth, and 250 MWth of power output, the criticality condition can be achieved if the reactor grade plutonium and MA content in loaded fuel are 7.56%, 6.76%, 6.56%, 6.56%, and 6.56%, respectively. The plutonium and MA utilization in small MSR results in the hardening of neutron spectrum compared to that of the standard 25 MWth miniFUJI MSR with Th-233 U fuel. Furthermore, the neutron spectra become harder with the augmenting of Pu and MA contents in loaded fuel as well as the increasing of the output power.

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Section: Methodsmentioning

confidence: 99%

Comparative Studies on Plutonium and Minor Actinides Utilization in Small Molten Salt Reactors with Various Powers and Core Sizes

et al. 2015

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“…Two main regions of FBR are designed for optimal fuel utilization from view point of fuel driver usage and for fuel breeding purpose which adopts 800 days as operation length per cycle for 4 fuel batches basis of Japan sodium fast reactor (JSFR) design as a basic core design analysis [11]. Optimization of reactor core analysis for space dependence and time dependence analysis has been done by adopting some established coupled computer code systems such as CITATION code, JOINT and SLAROM codes [12][13][14][15] which also produce some output datas for fuel composition behavior to analysis all actnide composition during reactor operation and its effect for reactor performance of FBR design and material barrier analysis based on plutonium composition.…”

Section: Analysis and Methodsmentioning

confidence: 99%

Analysis on Proliferation Resistance Factor and Fuel Breeding Capability Based on Even Mass Plutonium Isotope Compositions

et al. 2015

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Actinide recycling options of spent nuclear fuel are estimated to have many advantages such as reducing nuclear waste, increasing nuclear fuel utilization and improving proliferation resistance aspect of nuclear fuel. Composition of spent nuclear fuel consists of heavy nuclides and fission products including uranium and plutonium as main actinide. In addition, it is also consist of some minor actinides (MA) which can be recycled to reduce the volume and radioactive intensity of long-lived radioactive waste and in the same time to increase level of proliferation resistance. In general, some MAs can be utilized as useful material such as increasing nuclear fuel production and producing some even mass plutonium such as Pu-238, Pu-240 and Pu-242 as a control material for protected plutonium production. In this paper, identification of even mass plutonium isotopes production for the application of fuel breeding capability and proliferation resistant aspect of isotopic plutonium production was investigated. A large fast breeder reactor (FBR) model is adopted by using 800 days operation for one cycling operation for 4 fuel batches systems for optimization and analyses. The analysis of loading MA materials into the FBR core regions shows the Pu-238 production is increasing significantly as well as additional production of other plutonium isotopes. Isotopic Plutonium production of Pu-238 as main product can be estimated from fuel conversion process of MA in the reactor. This fuel conversion phenomena of Pu-238 can be identified as fertile material for fuel breeding capability, that Pu-238 can be converted to Pu-239 by neutron capture which is similar process also for fuel conversion of Pu-240 to Pu-241 as fissile materials. In addition, isotopic production of Pu-239 from converted Pu-238 as additional fuel which can be added with some quantities of Pu-239 production from converted U-238 for extending reactor operation. Production of Pu-238 in the reactor are used for categorizing some level of protected plutonium productions which can increase isotopic barrier level of plutonium in term of decay heat (DH) as well as Pu-240 as additional even mass contributor.

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“…The neutronics evaluation has been performed using the Monte-Carlo neutron transport calculation code MCNP-4C [8] and nuclear data library JENDL 3.2 [9]. As shown in Fig.…”

Section: Neutronics Calculationmentioning

confidence: 99%

Development of three-dimensional neutronics calculation system for design studies on helical reactor FFHR

Tanaka¹,

Sagara²,

Muroga³

et al. 2006

Fusion Engineering and Design

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Japanese Evaluated Nuclear Data Library Version 3 Revision-2: JENDL-3.2

Cited by 427 publications

References 24 publications

Comparative Studies on Plutonium and Minor Actinides Utilization in Small Molten Salt Reactors with Various Powers and Core Sizes

Comparative Studies on Plutonium and Minor Actinides Utilization in Small Molten Salt Reactors with Various Powers and Core Sizes

Analysis on Proliferation Resistance Factor and Fuel Breeding Capability Based on Even Mass Plutonium Isotope Compositions

Development of three-dimensional neutronics calculation system for design studies on helical reactor FFHR

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