SFCOMPO-2.0; An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

Michel-Sendis, F.*; Gauld, I.*; Martinez, J. S.*; Alejano, C.*; Bossant, M.*; Boulanger, D.*; Cabellos, O.*; Chrapciak, V.*; Conde, J.*; Fast, I.*; et al.

Annals of Nuclear Energy, 110, p.779 - 788, 2017/12

https://doi.org/10.1016/j.anucene.2017.07.022

Study on improvement of reactor physics analysis method for FBRs with various core concept (2)

Takeda, Toshikazu*; Tagawa, Akihiro; *; Kitada, Takanori*; *

JNC TJ9400 2001-009, 239 Pages, 2001/02

JNC-TJ9400-2001-009.pdf:8.71MB

Investigation was made on the following three themes as a part of the improvement of reactor physics analysis method for FBR with various core concepts. [Part 1: Investigations on Improvement of Neutron Spectrum Evaluation by the Use of Co-variance Matrices and Bias Corrections] In order to improve the neutron spectrum unfolding method used in the experimental fast reactor JOYO, investigation was made on the bias corrections to the initial neutron spectrum and error evaluation of nuclear data with the co-variance matrices. The error estimation was done by accumulating each bias correction factor and the co-variance matrix. It was concluded that the accumulated error for the initial neutron spectrum is relatively small, and a considerable improvement was achieved by the use of bias corrections. [Part 2: Evaluation of Neutron Streaming in Gas Cooled Fast Reactors by the use of Monte Carlo Method] As a part of investigations on the evaluation of the anisotropic diffusion coefficients for gas cooled fast reactors, a new tally function was added to a Monte Carlo code so that the neutron streaming can be calculated with heterogeneous core configurations. It was found that the neutron streaming becomes larger when the heterogeneous model was used. The tendency was more distinct in lower energy range. The same types of comparison was also done for the difference of core calculation models and the transport/diffusion theory. The final result shows that the transport/diffusion error has positive values in higher energy range, and the heterogeneous/homogeneous error has negative values in lower energy range. [Part 3: Investigation on the Calculation Method for Nuclear Converters with Neutron Moderators] A new calculation system which can deals with the target assemblies with neutron moderators was proposed. This concept has been investigated as a device to achieve high conversion rate for long life fission products. It was concluded that the characteristics method is ideal, wh

Improvement of numerical analysis method for FBR core characteristics (IV)

Takeda, Toshikazu*; *; Kitada, Takanori*

JNC TJ9400 99-002, 171 Pages, 1999/03

JNC-TJ9400-99-002.pdf:4.44MB

Investigation was made on the following three themes as a part of the improvement of numerical analysis method for FBR core characteristics. [Part 1: Improvement of Reaction Rate Calculation Method in Blanket Region.] A method to calculate multiband parameters directly from the precise energy structure of a cross section was established. This method can treat the precise neutron balance equation including the inter-band scattering. The procedure to treat the multiband effect with the conventional Sn code by the use of direction-dependent microscopic cross sections is shown. This procedure was applied to reaction rate distribution analysis on MONJU. As the result, the reaction rates increased in the blanket regions: the maximum increase was 1O% for U-238 cap, 12% for Pu-239 fis, 12% for U-235 fis, and 1% for U-238 fis. It became clear that the effect of the use of direction-dependent microscopic cross sections is small; thus most of the effect can be attributed to the change of microscopic cross section itself. [Part 2: Improvement of Reactivity Calculation by the Use of Monte Carlo Perturbation Method.] Applicability of the two Monte Carlo perturbation methods; the correlated sampling method and the derivative operator sampling method, to the perturbation problems with large change in material densities was investigated through numerical calculations. The result shows the fact that the present derivative operator sampling method, which treats only the first order term, is insufficient and higher order terms should be considered for such a large perturbation. Investigation was made on the new method which can treat the effect of adjoint flux change in energy and space distribution due to the perturbation, and a new formulation based on the exactperturbation was derived. [Part 3: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy.] A comparison was made between the two methods to improve the accuracy ...

Accuracy of MA transmutation calculation with MA specimens irradiated in the experimental fast reactor "JOYO"

*; Takeda, Toshikazu*; ; Kitada, Takanori; Aoyama, Takafumi

PNC TN9410 96-265, 64 Pages, 1996/05

PNC-TN9410-96-265.pdf:3.09MB

The accuracy to calculate the transmutation of minor actinides (MAs) was evaluated using MA specimens irradiated in the Experimental Fast Reactor JOYO. The Np transmutation characteristics were analyzed in this study. The transmutation ratio of Np was calculated by the one point burnup calculation code "ORIGEN2." The neutron fluxes which were input to "ORIGEN2" were calculated with the JOYO core management code system "MAGI" and "CITATION", both are based on three dimensional diffusion theory with 7 energy group. The continuous energy Monte-Carlo code "MVP" was introduced in order to evaluate the detail flux distribution inside the subassembly. The neutron flux obtained by a radiometric dose analysis without Np dosimeter was also used for the transmutation calculation. By comparing the flux distributions by MVP with and without the geometric heterogeneity in a subassembly, it was found that there had little effect on the calculation of Np transmutation near the core center of JOYO Mark-II. Calculated Np transmutation ratio with neutron fluxes by the MAGI and CITATION were 30% larger than that obtained from the -ray spectrum measurement, whereas the calculated value with the adjusted neutron flux by a radiometric dose analysis was about 1220% larger than the measured value. It is apparently due to an overestimation of the neutron fluxes by the MAGI and CITATION. As it was also found that there were large uncertainties in the adjusted neutron flux and the measured Np transmutation ratio, further investigation is required. As a result of this study, the points to be considered were found to improve the accuracy of the analysis.

A Feasibility study on MA loading in fast reactor cores

PNC TN9410 95-162, 121 Pages, 1995/06

PNC-TN9410-95-162.pdf:5.22MB

As one of the efforts to investigate the possibility of realizing an advanced fuel recycling system to recycle MA, a feasibility study is carried on upon a modified PUREX process which is based on the MOX fuel cycle. Although numerous studies have been made so far on the MA burning options in fast reactors, little effort has been devoted to apply the actual loading method of MA or to the proposal of taking measures against the presumable problems when such method is taken. In this study, which is based on design, an attempt to select a recycling process that satisfies all the design requirements and optimization through the evaluations on the merits and demerits in the overall recycle process is done. The selected option enables high performance in MA burning with low impact on the reactor operation or recycle facilities, in which Am is loaded with a non-U diluent in a target pin and placed in the fuel assembly with driver fuel pins. The feasibility study on the selected loading method considers the thermal criteria of the target pin, decay heat from spent fuel, and the impact on core performance. The selected option satisfies all the specified design criteria and at the same time high performance in MA burning.

PNC's Results on the Metal-Fueled Fast Reactor Benchmarks

Oki, Shigeo; Yamamoto, Toshihisa

PNC TN9410 95-001, 54 Pages, 1994/12

PNC-TN9410-95-001.pdf:1.39MB

WPPR(Working Party on Physics of Plutonium Recycling) has been organized in Nuclear Science Committee of OECD/NEA since November l992. More than ten advanced countries (France, United Kingdom, Germany, Russia, United States, Canada, Japan, etc.) participate in this working party. An aim of WPPR is to clarify some physical issues related to the technology for recycle of plutonium. To evaluate different scenarios for the use of plutonium, international benchmarks were developed for various types of reactors (MOX-fueled fast reactor, metal-fueled fast reactor, PWR and advanced converter). Among these, we contributed to the metal-fueled fast reactor benchmarks. In this report, our calculated results are summarized with all the information required. Each result is listed independently in a table according to the sequence indicated in the benchmark proposal, NEA/NSC/DOC(93)24.

None

;

Nihon Genshiryoku Gakkai-Shi, 36(11), p.1031 - 1038, 1994/11

https://doi.org/10.3327/jaesj.36.1031

None

None

; ; ; ; Aoyama, Takafumi; Uto, Nariaki;

PNC TN9410 93-024, 75 Pages, 1993/01

PNC-TN9410-93-024.pdf:6.54MB

None

JUPITER-III experiment analyses (III)

*; *; *; 7 of others*

PNC TN2410 89-003, 349 Pages, 1989/03

PNC-TN2410-89-003.pdf:8.55MB

The works of JUPITER sub-working group in FT'88 are summarized in this report. JUPITRR-III program is the co-operative research program between PNC and DOE, using ZPPR in ANL-W. The JUPITER sab-working group is organized by the reactor physics group of Reactor Research and Development Project, PNC, for planning and analysis of the experiment in JUPITER-III program. The first half of the JUPITER-III program, the ZPPR-17 program, was physics benchmark experiments to study the neutronic behavior of a 650 MWe-size, axially heterogeneous LMR core. The second half of the program, the ZPPR-18 program, was 1000 MWe-size two-zone homogeneous cores. ZPPR-17 was analysed almost completely, but ZPPR-18 analysis is being prepared. The main results of this year are as follows. (1)The following conclusions were obtained from the ZPPR-17 analysis. (i)In addition to the standard analysis model, the center-line and multi-drawer models were used. The effects on criticality were +0.17% for the former model and +0.080.09% for the latter model. The C/E values were almost the same as those for ZPPR-9 and ZPPR-13A. (ii)The C/E values were about 0.9 at the core center, and became about 5% higher at the core edge. The similar radial variations of C/E were observed in reaction rate distributions. The magnitude of 510% in the mispredictions are consistent with the previous values obtained in the analysis of the homogeneous core. (iii)The C/E values of sodium void reactivities were about 1.21.5 for the core regions, and about 0.8 for the internal blanket regions. The C/E values of sample reactirities in the internal blanket region were smaller than those in the core region by about 20% for Pu sample and by about 10% for B sample. The same underprediction in the internal blanket region was also seen in ZPPR-13A. (2)An attempt was made for building a data base system for the JUPITER program using the ZPPR-17A experimental data as an example. Some ...

None

Donen Giho, 75 Pages, 1986/00