Important fission product nuclides identification method for simplified burnup chain construction

Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi; Ohoka, Yasunori; Ushio, Tadashi

doi:10.1080/00223131.2015.1032381

Cited by 22 publications

(8 citation statements)

References 4 publications

(2 reference statements)

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“…A simplified burnup chain consisting of 138 FP nuclides, which are chosen by specific algorithm to reproduce number densities of some important nuclides after depletion [12], is used. Fission yields data and radioactive decay data of FP nuclides are taken from an updated version of JENDL FP yields data file-2011(JENDL/FPY-2011) and FP decay data file-2011(JENDL/FPD-2011) [13], in which nuclear data of over 1,300 FP nuclides are evaluated.…”

Section: Numerical Proceduresmentioning

confidence: 99%

“…Reference values of variance are obtained from sensitivity profiles calculated with an original detail chain and covariance data given in the evaluated nuclear data. A simplified chain consisting of 138 FP nuclides has been proposed so as to reproduce number densities of 33 FP nuclides, which have relatively large impact on reactivity [12], so number densities of these 33 FP nuclides are concerned in this verification calculation. Three sets of covariance matrices of a simplified chain are generated from a hundred, a thousand, and ten thousands samples.…”

Section: Verification Of Covariance Matrices For a Simplified Burnup mentioning

confidence: 99%

“…In our procedure of simplified burnup chain construction, unimportant nuclides, which are neglected in a simplified chain, are identified by observing adjoint number densities (or important functions) of these nuclides to target quantity. However, antimony-131 is neglected in the present simplified chain because of similarity in adjoint number density between antimony-131 and tellurium-131, which is different criteria in our important nuclides identification process[12]. We do not yet suffi-…”

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confidence: 94%

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Uncertainty quantification of neutron multiplication factors of light water reactor fuels during depletion

Chiba

Okumura

2018

Journal of Nuclear Science and Technology

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Nuclear data-induced uncertainties of infinite neutron multiplication factors (k ∞) during fuel depletion are quantified in a single cell and a 3×3 multi-cell including burnable absorbers. Uncertainties of reaction cross sections, fission yields, decay half-lives and decay branching ratios provided in the JENDL libraries are taken into account. 100% uncertainties are assumed to nuclear data to which uncertainty information are not provided in JENDL. Uncertainties propagation calculations are carried out with the adjoint-based procedure, and required sensitivity profiles of k ∞ with respect to these nuclear data are efficiently calculated by the depletion perturbation theory. Covariance matrices for fission yields and decay data in a simplified burnup chain are successfully generated by the stochastic-based procedure. k ∞ uncertainties of about 0.6% during fuel depletion are obtained, and it is shown that actinoids reaction cross sections are dominant contributors. Nuclide-wise decomposition of the uncertainties and observation of component-wise sensitivity profiles provide physical interpretations. By virtue of the adjoint-based procedure, several parametric surveys are also conducted. Contributions of uncertainties in fission products (FP) nuclides are quantified, and important nuclides and energy ranges are identified for further evaluation of nuclear data of FP nuclides. Effect of cooling period on k ∞ uncertainties is also discussed.

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Section: Numerical Proceduresmentioning

confidence: 99%

Section: Verification Of Covariance Matrices For a Simplified Burnup mentioning

confidence: 99%

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confidence: 94%

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Uncertainty quantification of neutron multiplication factors of light water reactor fuels during depletion

Chiba

Okumura

2018

Journal of Nuclear Science and Technology

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“…The isotopic compositions in five fuel-gadolinium pin radial zones for burnup point 40 MWd/kgHM ( 235 U and 239 Pu) and for burnup point 2 MWd/kgHM ( 155 Gd and 157 Gd) are presented in Figs. [16][17][18][19]. The radial distributions of the 235 U, 239 Pu and 155 Gd calculated by MCNP5-ORIGEN2 coupling scheme showed good agreement with the benchmark results.…”

Section: Isotopic Composition Versus Fuel Volume Radiusmentioning

confidence: 70%

Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies

Nguyen¹,

Nguyen²,

Tran³

et al. 2016

Nucl. Sci. and Tech.

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The paper aims to develop an MCNP5-ORIGEN2 coupling scheme for burnup calculation. Specifically, the Monte Carlo neutron transport code (MCNP5) and the nuclides depletion and decay calculation code (ORIGEN2) are combined by data processing and linking files written in the PERL programming language. The validity and applicability of the developed coupling scheme are tested through predicting the neutronic and isotopic behavior of the “VVER-1000 LEU Assembly Computational Benchmark”. The MCNP5-ORIGEN2 coupling results showed a good agreement with the k-inf benchmark values within 600 pcm during the entire burnup history. In addition, the differences of isotopes concentration at the end of the burnup (40 MWd/kgHM) when compared with benchmark values were reasonable and generally within 6.5%. The developed coupling scheme also considered the shielding effect due to gadolinium isotopes and simulated well the depletion of isotopes as a function of the radial position in gadolinium bearing fuel rods.

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“…Neutron flux is normalized with constant line power of 179 W/cm per one fuel pin in average. Anuclide depletion chain consisting of 138 fission product nuclides, which is optimized for light water reactor reactivity calculations[13], is employed. Length of one depletion step is 1.6 GWD/t while finer depletion steps are used at the beginning and end of the depletion.…”

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confidence: 99%

Perturbation theory for nuclear fuel depletion calculations with predictor–corrector method

Chiba

2017

Journal of Nuclear Science and Technology

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The perturbation theory for nuclear fuel depletion calculations with the predictorcorrector method is derived. This theory is implemented to a reactor physics code system CBZ, and the theory itself and its implementation are numerically verified. Sensitivities of nuclide number densities after fuel depletion with respect to nuclear data calculated with this theory are compared with reference sensitivities calculated by numerical differentiation, and good agreements are obtained. Importance of accurate angle integration on product of neutron flux and generalized adjoint neutron flux is also pointed out. Sensitivities in a 3×3 multicell system including a gadolinium-bearing fuel pin are calculated, and it is demonstrated that the derived theory yields accurate sensitivities even if coarse depletion time step division is adopted. The present work drastically increases the applicability of the depletion perturbation theory to actual problems.

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Important fission product nuclides identification method for simplified burnup chain construction

Cited by 22 publications

References 4 publications

Uncertainty quantification of neutron multiplication factors of light water reactor fuels during depletion

Uncertainty quantification of neutron multiplication factors of light water reactor fuels during depletion

Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies

Perturbation theory for nuclear fuel depletion calculations with predictor–corrector method

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