Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment

Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 61(1), p.31 - 43, 2024/01

https://doi.org/10.1080/00223131.2023.2244512

This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g., reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.

Development of a standard data base for FBR core nuclear design, 12; Analysis of FCA X-1 experiments and consistency evaluation using cross-section adjustment

Yokoyama, Kenji*; Numata, Kazuyuki*; Ishikawa, Makoto*; Oigawa, Hiroyuki; Iijima, Susumu

JNC-TY9400 2000-006, 168 Pages, 2000/04

no abstracts in English

Analysis of measurements for a Uranium-free core experiment at the BFS-2 critical assembly

Hunter

JNC TN9400 99-049, 74 Pages, 1999/04

JNC-TN9400-99-049.pdf:2.03MB

This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2d) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of keff was 1.1%k/k higher than the measured value, Na void worth C/E values were 1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes, though the efect should be investigated in any future experiments.) several sample worth values were small compared with calculational uncertaint

Development of a standard data base for FBR core nuclear design(II); JUPITER-I experlmental data book

*

PNC TN9410 93-010, 502 Pages, 1992/12

PNC-TN9410-93-010.pdf:17.39MB

The present report compiles the experimental data of JUPITER phase-I, which was a joint research program between U.S.DOE and PNC of Japan, using the ZPPR facility at ANL-Idaho in 1978 to 1979. The JUPITER-I experiment was a series of critical experiments for conventional two-zone homogeneous cores of 600 to 800 MWe-class LMFBR, including seven experimental cores The nuclear characteristics recorded here include criticality, control rod reactivity, reaction rate distribution, sodium void reactivity, sample reactivity, Doppler reactivity, gamma heating and neutron spectrum. (1)ZPPR-9 : two-region cylindrical clean core with volume of app. 4,600 liters, (2)ZPPR-10A : hexagonal engineering-mockup core with 19 cotrol-rod positions(CRPs), (3)ZPPR-10B : changes seven CRPs to control rods(CRs) from ZPPR-10A, (4)ZPPR-10C : volume of app. 6,200 liters with similar core arrangement to ZPPR-10A, (5)ZPPR-10D : 31 CRPs with the same volume as ZPPR-10C, (6)ZPPR-10D/1 : changes the central CRP to a CR from ZPPR-10D, and, (7)ZPPR-10D/2 : changes seven CRPs to CRs from ZPPR-10D. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The detail of experimental data is thoroughly recorded here so as to re-analyze these experiments in future. In addition, these experimental data are installed in the computer system at OEC for convenience of analytical code input.