Study of Nuclear Decay Data Contribution to Uncertainties in Heat Load Estimations for Spent Fuel Pools

At the Paul Scherrer Institut (PSI), a methodology for nuclear data uncertainty propagation in CASMO-5M (C5M) assembly calculations is under development. This paper presents a preliminary application of this methodology to C5M decay heat calculations. Applying a stochastic sampling method, nuclear decay data uncertainties are first propagated for the cooling phase only. Thereafter, the uncertainty propagation is enlarged to gradually account for cross-section as well as fission yield uncertainties during the depletion phase. On that basis, assembly heat load uncertainties as well as total uncertainty for the entire pool are quantified for cooling times up to one year. The relative contributions from the various types of nuclear data uncertainties are in this context also estimated.

Development and Application of Data Assimilation Methods in Reactor Physics

Daniel Jerôme Siefman

Simulations of nuclear reactor physics can disagree significantly from experimental evidence, even when the most accurate models are used. An important part of this bias from experiment is caused by nuclear data. The nuclear data have inherent uncertainties due to the way they are evaluated, which then propagate to nuclear reactor simulations. This creates a bias and an uncertainty in a predicted reactor parameter like \keff~or the composition of spent fuel. This thesis focuses on data assimilation techniques to ameliorate the effects of nuclear data. Data assimilation takes integral experiments and assimilates them in a Bayesian way to improve simulations. It can also be used to find trends and areas needing improvement in evaluated nuclear data. The research focuses on advancing the data assimilation theory and knowledge used in reactor physics, especially on techniques that require stochastic sampling of the nuclear data. Furthermore, the research takes advantage of rich experimental data available from the Proteus research reactor at the Paul Scherrer Institute. The thesis showed, for the first time, that two methods based on stochastic sampling (called MOCABA and BMC) gave equivalent results to each other and to the traditional method called GLLS. This was corroborated with two independent studies that used different experiments, neutron transport codes, nuclear data, and processing codes. The first study used the JEZEBEL-Pu239 benchmark, the Serpent2 neutron transport code, and NUSS. The second study used reactivity experiments from the LWR-Phase II experiments at Proteus, CASMO-5 for neutron transport, and SHARK-X. While using Serpent2, several questions pertaining to the stochastic uncertainty of its sensitivity coefficients arose. To address these, a new method called eXtended GLLS, or xGLLS, was proposed and tested in the thesis. xGLLS showed that the uncertainties associated with sensitivity coefficients have a negligible effect on the data assimilation as long as the calculated integral parameters themselves were converged. The final study focused on adjusting the fission yields and covariances made by the GEF code with post-irradiation examination experiments from Proteus. The adjustment improved the accuracy of predicted nuclide concentrations in spent fuel and improved the agreement between the GEF fission yields and those of ENDF/B-VIII.0 and JEFF3.3.

EPFL2019

Determination of the activity inventory and associated uncertainty quantification for the CROCUS zero power research reactor

Mathieu Hursin, DongHyuk Lee, Andreas Pautz

The paper describes the source term estimation of CROCUS, the zero power research reactor of EPFL, to be used for dispersion analysis under accidental conditions. To fulfil regulatory requirements, the source term of the CROCUS fuel is estimated through Monte Carlo simulations supplemented by uncertainty quantification, both obtained from the Monte Carlo code MCS developed at UNIST. Even though the depletion capabilities of MCS were pre-existing to this work, no verification has been documented so far. A comparison of MCS and SERPENT results for the determination of the activity inventory for the CROCUS fuel is presented; both codes agree within the 1% for the major isotopes contributing to both inhaled and ingested doses. The source term of CROCUS is calculated under a postulated accident. Eight isotopes, 90Sr, 91Y, 131I, 137Cs, 140Ba, 140La, 144Ce, and 239Pu produce the largest contributions to the effective dose to the public. For uncertainty quantification, nuclear data uncertainties, specifically cross sections and fission yields are considered. Three stochastic sampling methods are implemented in the MCS code namely, TMC, fast-TMC and fast-GRS methods. The performances of the two fast methods have been analyzed when applied to the CROCUS reactor. All three methods produce consistent uncertainty estimates and the two fast methods showed a reduced computational cost compared to the original TMC method. The fission yield uncertainty is the leading factor for the determination of uncertainty of the activity for 90Sr, 91Y, 131I, 137Cs, 140Ba, 140La and 144Ce isotopes. On the other hand, the cross section uncertainty is the leading factor for the uncertainty of 239Pu activity. Finally, the modeling of the irradiation history for CROCUS is simplified to reduce the computational cost. It is demonstrated that the activities for the short lived isotopes (131I and 144Ce) are very sensitive to the irradiation history specifications. Nonetheless, the irradiation history used for the determination of the nominal fuel inventory activity is conservative as it overestimates the activity of the short lived isotopes.

2019

Determination of the activity inventory and associated uncertainty quantification for the CROCUS zero power research reactor

Mathieu Hursin, DongHyuk Lee, Andreas Pautz

2019

Development and Application of Data Assimilation Methods in Reactor Physics

Daniel Jerôme Siefman

EPFL2019