R-Matrix Analysis of the<sup>240</sup>Pu Neutron Cross Sections in the Thermal to 5700-eV Energy Range

Bouland, O.; Derrien, H.; Larson, N. M.; Leal, L.C.

doi:10.13182/nse127-105

Cited by 16 publications

(29 citation statements)

References 13 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Despite the evaluation works performed on these isotopes for the JEFF-3.1.1 library, we still observe systematic overestimation of the experimental k eff by results of calculations that increases with the Pu-ageing, which defines the 241 Am build up from 241 Pu beta decay with time. For the reference configuration of the MISTRAL-4 experiments, the bias on the calculation-to-experimental ratios reaches +735 (8) pcm. Such biases were recently reduced with the latest version of the JEFF library, namely JEFF-3.2.…”

Section: Introductionmentioning

confidence: 94%

Improved Mixed Oxide Fuel Calculations with the Evaluated Nuclear Data Library JEFF-3.2

Noguère

Bernard

Blaise

et al. 2016

Nuclear Science and Engineering

Self Cite

View full text Add to dashboard Cite

An overestimation of the k eff values for MOX fuels was identified with Monte-Carlo (TRIPOLI4) and deterministic (APOLLO2) calculations based on the Evaluated Nuclear Data Library JEFF. The overestimation becomes sizeable with the Pu-ageing, reaching a reactivity change of ∆ρ ≃ +700 pcm for integral measurements carried out with MOX fuel containing a large amount of Americium. This bias was observed on various critical configurations performed in the zero power reactor EOLE of the CEA Cadarache, France. The present work focuses on the improvements achieved with the new 239 Pu and 241 Am Evaluated Nuclear Data Files available in the latest version of the JEFF library (JEFF-3.2). The Resolved Resonance Range of the plutonium evaluation was re-evaluated at ORNL (Oak Ridge National Laboratory) with the SAMMY code in collaboration with the CEA Cadarache. The resonance parameters of the Americium evaluation were obtained with the REFIT code in collaboration with the research institutes IRMM

show abstract

Section: Introductionmentioning

confidence: 94%

Improved Mixed Oxide Fuel Calculations with the Evaluated Nuclear Data Library JEFF-3.2

Noguère

Bernard

Blaise

et al. 2016

Nuclear Science and Engineering

Self Cite

View full text Add to dashboard Cite

show abstract

“…The details of this methodology are given in the SAMMY manual [41] and have been applied in previous evaluations [40]. (ix) A normalization factor scaling the capture yield and a constant background have been fitted in each of the intervals in which the resonance region was divided.…”

Section: A Resonance Analysis Strategy With Sammymentioning

confidence: 99%

“…These result from the propagation of the uncertainties and the correlation between resonance parameters through the sequential fit of the capture and transmission data (see Ref. [40] …”

Section: -7mentioning

confidence: 99%

Measurement and resonance analysis of the237Np neutron capture cross section

Guerrero¹,

Cano‐Ott²,

Mendoza³

et al. 2012

Phys. Rev. C

View full text Add to dashboard Cite

The neutron capture cross section of 237 Np was measured between 0.7 and 500 eV at the CERN n_TOF facility using the 4π BaF 2 Total Absorption Calorimeter. The experimental capture yield was extracted minimizing all the systematic uncertainties and was analyzed together with the most reliable transmission data available using the SAMMY code. The result is a complete set of individual as well as average resonance parameters [D 0 = 0.56(2) eV, γ = 40.9(18) meV, 10 4 S 0 = 0.98(6), R = 9.8 (6) (2002)]. The results indicate that a new evaluation combining the present capture data with reliable transmission data would allow reaching an accuracy better than 4%, in line with the uncertainty requirements of the nuclear data community for the design and operation of current and future nuclear devices.

show abstract

“…The measured capture yields of 237 Np and 240 Pu have been analysed using the SAMMY code [9], following the iterative SAMMY sequential fit method in the resolved resonance region (RRR) [10,11]. The capture yields are analyzed together with transmission, using the resulting resonance parameters and covariance matrix of the fit to one data set as input for the next one.…”

Section: Cross Section Analysismentioning

confidence: 99%

“…The n TOF capture data are analyzed together with the transmission data of Kolar and Böckoff [17] (GELINA, 1968), also included in the work of Bouland et al [10] and the evaluated data files. The 240 Pu level spacing is D ∼ 12 eV and the resonances are well resolved up to several keV.…”

Section: Pu Cross Sectionmentioning

confidence: 99%

The neutron capture cross sections of ²³⁷Np(n,γ) and ²⁴⁰Pu(n,γ) and its relevance in the transmutation of nuclear waste

Guerrero

Abbondanno

Aerts

et al. 2007

Nd2007

View full text Add to dashboard Cite

Abstract. Neutron capture cross sections of actinides are of great relevance for the Transmutation of Nuclear Waste in Accelerator Driven Systems (ADS) and Generation-IV reactors. The neutron capture cross sections of 237 Np and 240 Pu were measured at the n TOF facility with a Total Absorption Calorimeter. The data have been analyzed with the SAMMY code. The corresponding covariance matrices have been generated. The final cross sections are presented and compared to the previously existing ones.

show abstract

R-Matrix Analysis of the²⁴⁰Pu Neutron Cross Sections in the Thermal to 5700-eV Energy Range

Cited by 16 publications

References 13 publications

Improved Mixed Oxide Fuel Calculations with the Evaluated Nuclear Data Library JEFF-3.2

Improved Mixed Oxide Fuel Calculations with the Evaluated Nuclear Data Library JEFF-3.2

Measurement and resonance analysis of the237Np neutron capture cross section

The neutron capture cross sections of ²³⁷Np(n,γ) and ²⁴⁰Pu(n,γ) and its relevance in the transmutation of nuclear waste

Contact Info

Product

Resources

About

R-Matrix Analysis of the240Pu Neutron Cross Sections in the Thermal to 5700-eV Energy Range

Cited by 16 publications

References 13 publications

Improved Mixed Oxide Fuel Calculations with the Evaluated Nuclear Data Library JEFF-3.2

Improved Mixed Oxide Fuel Calculations with the Evaluated Nuclear Data Library JEFF-3.2

Measurement and resonance analysis of the237Np neutron capture cross section

The neutron capture cross sections of 237Np(n,γ) and 240Pu(n,γ) and its relevance in the transmutation of nuclear waste

Contact Info

Product

Resources

About

R-Matrix Analysis of the²⁴⁰Pu Neutron Cross Sections in the Thermal to 5700-eV Energy Range

The neutron capture cross sections of ²³⁷Np(n,γ) and ²⁴⁰Pu(n,γ) and its relevance in the transmutation of nuclear waste