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RISS 인기검색어
- 검색키워드
- 저자 : D. Rochman (검색결과 7 건)
- 무료
- 기관 내 무료
- 유료
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Nuclear Data Uncertainty Propagation for a Sodium Fast Reactor
D. Rochman,A. J. Koning,D. F. Dacruz,S. C. van der Marck 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23
Nuclear data uncertainties are propagated from basic theory to a full core model of the Kalimer-600 Korean type Sodium Fast Reactor (SFR) using the TALYS nuclear code system, the "Total Monte Carlo" (TMC) approach and perturbation methods developed at NRG. Nuclear data uncertainties from sodium, iron, and some actinides will be presented together with their impact on parameters such as the sodium void coefficient and k<sub>eff</sub>. One of the advantages of the TMC method is that it avoids approximations used in perturbation theories, by applying an exact uncertainty propagation approach. Additionally, full nuclear data uncertainties (cross sections, nu-bar as well as single and double differential data) can be considered.
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Exact Nuclear Data Uncertainty Propagation for Fusion Design
D. Rochman,A. J. Koning,S. C. van der Marck 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23
Recently, we have presented an exact method (now called ``Total Monte Carlo'') to propagate uncertainties of fundamental nuclear physics experiments, models and parameters to large and complicated nuclear systems. We now show that such exact uncertainty calculations are directly relevant to the optimal and safe design of fusion systems by applying this methodology to a series of fusion shielding benchmarks, namely those connected to the Oktavian, FNS and LLNL experiments. Uncertainties on neutron and gamma leakage fluxes for shielding benchmarks are obtained. Uncertainties for cross sections, angular distributions, single- and double-differential emission spectra, and gamma-ray production cross sections are considered.
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Nuclear Data Uncertainty Propagation for a Typical PWR Fuel Assembly with Burnup
D. ROCHMAN,C.M. SCIOLLA 한국원자력학회 2014 Nuclear Engineering and Technology Vol.46 No.3
The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECDNuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The “Fast Total Monte Carlo”method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on k∞, reaction rates,two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections forthe actinides and fission products, fission yields and thermal scattering data.
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Nuclear Data Uncertainty Propagation: Total Monte Carlo vs. Covariances
D. Rochman,A. J. Koning,S. C. van der Marck,A. Hogenbirk,D. van Veen 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23
Two distinct methods of propagation for basic nuclear data uncertainties to large scale systems will be presented and compared. The "Total Monte Carlo" method is using a statistical ensemble of nuclear data libraries randomly generated by means of a Monte Carlo approach with the TALYS system. These libraries are then directly used in a large number of reactor calculations (for instance with MCNP) after which the exact probability distribution for the reactor parameter is obtained. The second method makes use of available covariance files and can be done in a single reactor calculation (by using the perturbation method). In this exercise, both methods are using consistent sets of data files, which implies that covariance files used in the second method are directly obtained from the randomly generated nuclear data libraries from the first method. This is a unique and straightforward comparison allowing to directly apprehend advantages and drawbacks of each method. Comparisons for different reactions and criticality-safety benchmarks from ^(19)F to actinides will be presented. We can thus conclude whether current methods for using covariance data are good enough or not.
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NUCLEAR DATA UNCERTAINTY PROPAGATION FOR A TYPICAL PWR FUEL ASSEMBLY WITH BURNUP
Rochman, D.,Sciolla, C.M. Korean Nuclear Society 2014 Nuclear Engineering and Technology Vol.46 No.3
The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The "Fast Total Monte Carlo" method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on $k_{\infty}$, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.
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Modern Nuclear Data Evaluation: Straight from Nuclear Physics to Applications
A. J. Koning,D. Rochman 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23
The nuclear data field comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation.A software system, built around the TALYS code, will be presented in which all of these essential nuclear data components are seamlessly integrated into one approach. This system brings back nuclear data evaluation to its essence: The starting information is no longer an ENDF-6 file, but a set of selected differential experiments, resonance parameters and a nuclear model input file.After this, computer power does the rest.The implications of this are unprecedented. A few are:\begin{itemize}\item Complete ENDF-6 nuclear data libraries, such as TENDL, including covariance matrices, for many isotopes, particles, energies, reaction channels and secondary quantities. All isotopic data files are mutually consistent and will soon rival those of the major world libraries.\item Exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach (``Total'' Monte Carlo).\end{itemize}
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A. Dokhane,A. Vasiliev,M. Hursin,D. Rochman,H. Ferroukhi 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.5
The aim of this paper is to assess the reliability and accuracy of the PSI standard method, used in manyprevious works, for the quantification of ND uncertainties in the SPERT-III RIA transient, by quantifyingthe discrepancy between the actual inserted reactivity and the original static reactivity worth and theirassociated uncertainties. The assessment has shown that the inherent S3K neutron source renormalization scheme, introduced before starting the transient, alters the original static reactivity worth of thetransient CR and reduces the associated uncertainty due to the ND perturbation. In order to overcomethese limitations, two additional methods have been developed based on CR adjustment. The comparative study performed between the three methods has showed clearly the high sensitivity of the obtained results to the selected approach and pointed out the importance of using the right procedure inorder to simulate correctly the effect of ND uncertainties on the overall parameters in a RIA transient. This study has proven that the approach that allows matching the original static reactivity worth andstarting the transient from criticality is the most reliable method since it conservatively preserves theeffect of the ND uncertainties on the inserted reactivity during a RIA transient.
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