Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

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Critical experiments were performed in the REBUS program on a core loaded with a test bundle including 16 irradiated BWR-type MOX rods of average burnup of 61 GWd/t. The experimental data were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 or JENDL-3.3. Biases in effective multiplication factors of the critical cores were 71.0%Dk for the diffusion calculations (JENDL-3.2), 70.3%Dk for the transport calculations (JENDL-3.3), and 0.2%Dk for the Monte Carlo calculations (JENDL-3.2). The measured core fission rate and co-activation rate distributions were generally well reproduced using the three types of calculations. The burnup reactivity determined using the measured water level reactivity coefficients was 72.41 + 0.08%Dk/kk', which also agreed with the results of the three type of calculations within the measurement and calculation errors. The most probable isotopic inventories in the irradiated MOX rods was tentatively obtained by using the ratios of the calculation to chemical assay data on a pellet sample, and the burnup reactivity was reanalyzed to split the calculation error into those due to the inventory and reactivity calculations. This approach showed that the inventory calculation error compensated the reactivity calculation error.

Section: Analysis Results Of Core Characteristicssupporting

confidence: 84%

Section: Monte Carlo Calculationmentioning

confidence: 70%

Section: Monte Carlo Calculationmentioning

confidence: 91%

“…The analysis methods applied to the experimental cores were basically the same as those described in Refs. [6,16].…”

Section: Calculation Methodsmentioning

confidence: 99%

See 2 more Smart Citations

Analysis of core physics experiments on irradiated BWR MOX fuel in the REBUS program

Yamamoto

Ando

Hayashi

et al. 2012

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“…In this study, the measurement data of the institutes A, B, and A were referred for the MOX, UO 2 -M, and UO 2 -L samples, respectively. This is because (1) the present authors have many experiences to analyze the isotopic composition data of the institute A [3,[9][10][11] for samples taken from MOX and UO 2 fuel assemblies and confirmed the characteristics of the measured data, (2) the institute B adopted a similar measurement technique to the institute A such as mass spectrometry, and (3) the data of the institute C often showed different trends from those of the other institutes.…”

Section: Isotopic Composition Measurements On Pwr Fuel In Malibu Programsupporting

confidence: 67%

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO₂fuels in the MALIBU program

Yamamoto¹,

Suzuki²,

Ando³

et al. 2012

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The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO 2 assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 6 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 6 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 6 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO 2 sample of 72 GWd/t.

Feedback on neutron capture cross sections of ²³⁸Pu and ²⁴¹Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

Yamamoto¹,

Sakai²

2015

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The atomic fractions of 238 Pu and 241 Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for 238 Pu and 241 Am. The thermal neutron capture cross sections of 238 Pu and 241 Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of 238 Pu on UO 2 and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of 238 Pu and 241 Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.