Amount of Nuclides Constituting PWR Spent Fuels

Nakahara, Y.; Suzuki, Toshio; Gunji, Katsubumi; Kohno, Nobuaki; Takano, Hideki; Yoshida, Zenko; Adachi, T.

doi:10.1524/ract.1990.50.3.141

Cited by 12 publications

(6 citation statements)

References 4 publications

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“…We also find a strong correlation between the two Pu isotopes, which can be explained by the fact that 239 Pu is a precursor to 241 Pu in one of three main isotope transformation chains within the reactor core [63]:…”

Section: Uncertainties In Fission Calculationsmentioning

confidence: 70%

Uncertainties in the anti-neutrino production at nuclear reactors

Djurcic

Detwiler

Piepke

et al. 2009

J. Phys. G: Nucl. Part. Phys.

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Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates inνe detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties, and their relevance to reactorνe experiments.

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Section: Uncertainties In Fission Calculationsmentioning

confidence: 70%

Uncertainties in the anti-neutrino production at nuclear reactors

Djurcic

Detwiler

Piepke

et al. 2009

J. Phys. G: Nucl. Part. Phys.

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“…Details of the experimental procedure and results are reported in Ref. (20). The following 6 burn-up calculations were carried out for the most irradiated sample whose exposure is 34.1 GWd/t.…”

Section: Irradiated Fuel Sample and Calculation Methodsmentioning

confidence: 99%

“…The irradiation history of the sample fuel is given in a literatureC 20 l, which is shown in Fig. 10.…”

Section: Irradiated Fuel Sample and Calculation Methodsmentioning

confidence: 99%

Validation of a Continuous-Energy Monte Carlo Burn-up Code MVP-BURN and Its Application to Analysis of Post Irradiation Experiment.

Okumura

Mori

Nakagawa

et al. 2000

J. Nucl. Sci. Technol.

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In order to confirm the reliability of a continuous-energy Monte Carlo burn-up calculation code MVP-BURN, it was applied to the burn-up benchmark problems for a high conversion LWR lattice and a BWR lattice with burnable poison rods. The results of MVP-BURN have shown good agreements with those of a deterministic code SRAC95 for burn-up changes of infinite neutron multiplication factor, conversion ratio, power distribution, and number densities of major fuel nuclides. Serious propagation of statistical errors along burn-up was not observed even in a highly heterogeneous lattice. MVP-BURN was applied to the analysis of a post irradiation experiment for a sample fuel irradiated up to 34.1 GWd/t, together with SRAC95 and SWAT. It was confirmed that the effect of statistical errors of MVP-BURN on a burned fuel composition was sufficiently small, and it could give a reference solution for other codes. In the analysis, the results of the three codes with JENDL-3.2 agreed with measured values within an error of 10% for most nuclides. However, large underestimation by about 20% was obse.rved for 238 Pu, 242 m Am and 244 Cm. It is probable that these discrepancies are a common problem for most current nuclear data files.

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“…After dissolution, radioactivities of 13'Cs in the diluted sample solutions were determined using standard radioactivity sources of of Spent Fuel 234U, 237Np, 2 3 6 p u , 238pu, 241Am, Z43Am, 242Cm and 152Eu. 3…”

Section: Analytical Methods")mentioning

confidence: 99%

“…In Table 3 shown are the irradiation histories of these five spent fuel samples. These values were obtained from the irradiation history of each fuel rod and the 3 Cutting positions (mm) of samples for three fuel rods used (sample name written in parenthesis) measured burnup, with considering the cutting position and the neutron flux distribution along the fuel rod based on the relative power distribution of the fuel rod.…”

Section: It Experimental 1 Pwr Spent Fuel Samplesmentioning

confidence: 99%

Comparison of Calculated Values with Measured Values on the Amount of TRU and FP Nuclides Accumulated in Gadolinium Bearing PWR Spent Fuels.

Adachi

Nakahara²,

Kohno³

et al. 1994

J. Nucl. Sci. Technol.

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Destructive analyses for five spent fuel samples taken from a Gd bearing fuel assembly were done. The measured amounts of actinides of zs4-zssU, zs7Np, 23*-z4zP~, z41~242m~z43Am and , and fission products of Is4Cs and 1 5 4 E~ were used for evaluating the accuracy of calculation made by CASMO-MICBURN and ORIGEN-2 codes. T h e effect of Gd on the neutron spectrum was taken into account in the CASMO-MICBURN calculation.The amounts of 235U, 239Pu and 241Pu calculated by CASMO-MICBURN agreed well with the observed values within about 3%. On the other hand, the amounts obtained from ORIGEN-2 calculation showed lower values than those observed, especially by -12% in average in 236ufor Gdz03-U02 fuel. 'The main cause of this large difference may be attributed to the effect of Gd on the neutron spectrum. The amounts of the other actinides by both calculation codes revealed no significant difference in nearly 10% except for 24zmAm, in which a large fluctuation among the samples was observed. About 10% difference between the measured values and the calculated values was also observed for Is4Cs, but the calculated values for IS4Eu showed a significant difference from measured values. 242,244Cm

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Amount of Nuclides Constituting PWR Spent Fuels

Abstract: Destructive analysis of nuclides in PWR spent fuels whose burnup rates were between 6.9 and 34.

Cited by 12 publications

References 4 publications

Uncertainties in the anti-neutrino production at nuclear reactors

Uncertainties in the anti-neutrino production at nuclear reactors

Validation of a Continuous-Energy Monte Carlo Burn-up Code MVP-BURN and Its Application to Analysis of Post Irradiation Experiment.

Comparison of Calculated Values with Measured Values on the Amount of TRU and FP Nuclides Accumulated in Gadolinium Bearing PWR Spent Fuels.

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