Reevaluation of experimental data and analysis on experimental fast reactor JOYO MK-I performance tests

Yokoyama, Kenji; Shono, Akira*; Ishikawa, Makoto

Nuclear Science and Engineering, 157(3), p.249 - 263, 2007/11

https://doi.org/10.13182/NSE06-109

Experimental data acquired in the experimental fast reactor JOYO MK-I performance tests in the late 1970s have been revaluated and analyzed with a nuclear analysis system for fast reactors used in Japan Atomic Energy Agency (JAEA). For the purpose of improving the prediction accuracy of nuclear characteristics, nominal values and uncertainties of the experimental data were revaluated by using knowledge obtained after the MK-I performance test and calculation results based on the latest reactor physics analysis methods. All the nominal values were corrected by using a formulation of control rod interaction effects proposed in the present paper, and all the possible uncertainty factors were evaluated and quantified. The analysis results agreed well with measured values within the experimental and nuclear-induced uncertainties for all the nuclear characteristics of the criticality, control rod worth sodium void reactivity, fuel replacement reactivity and isothermal temperature coefficient.

Reevaluation of Experimental Data and Analysis with the Latest Reactor Physics Calculation Method on Fast Experimental Reactor JOYO MK-I Perfomance Tests

Yokoyama, Kenji; Numata, Kazuyuki*; Shono, Akira; Ishikawa, Makoto

JNC TN9400 2005-024, 372 Pages, 2005/05

JNC-TN9400-2005-024.pdf:25.83MB

"JOYO" MK-I has a typical fast breeder reactor core which was fuelled by plutonium-uranium mixed oxide (MOX) and the fuel region was surrounded by blanket consisting of depleted uranium oxide. Since it has a simple core geometry without any other special irradiation subassembly, it is suitable for the reactor physics analysis. The experimental data acquired in the "JOYO" MK-I performance tests are analyzed with the latest analysis methods in order to resister the data in the IRPhE (International Reactor Physics Benchmark Experiments) project. For this analysis, nominal values and uncertainties of the experimental data were reevaluated by using knowledge obtained after the MK-I performance test and calculation results based on the latest analysis methods. As the nuclear characteristics for the analysis, we selected the criticality, the control rod worth, the isothermal temperature coefficient, the fuel replacement reactivity worth and the sodium void reactivity worth, which were measured in unburnt cores prior to the first power ascension to 50MWth. In this evaluation, not only the measurement uncertainty but also geometry uncertainty and composition uncertainty are considered. All the uncertainties are evaluated with a classification into the random and the systematic components, according as the guideline of IRPhE. The analysis of "JOYO" MK-I was previously carried out in the fiscal year of 1999. In this report, the analysis is totally carried out again with the revised group constants and the latest analysis codes such as the ultra-fine group lattice calculation code. In order for use in the benchmark problem preparation, the analytical results are shown with various correction factors that are estimated from the difference between analytical models. The uncertainties of analysis results are also evaluated by considering the correction factors. Furthermore, a consistency evaluation with the other existing critical experiment data is carried out by performing ...

Reduction of Cross-Section-Induced Errors of the BN-600 Hybrid Core Nuclear Parameters by Using BFS-62 Critical Experiment Data

Shono, Akira; Hazama, Taira; Ishikawa, Makoto; Manturov, G.*

Proceedings of International Conference on the Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004), 95315 Pages, 2004/00

The present paper provides evaluation results of predicted uncertainty on nuclear parameters on the BN-600 hybrid core, a feasible option for Russian surplus weapons plutonium disposition. Covariance of nuclear group constant, analysis error, and experimental error are considered to predict uncertainties of the hybrid core nuclear parameters by applying the nuclear group constant adjustment method. Analysis results of BFS mockup and other fast reactor core experiments were reflected in the evaluation.

Effects of Nuclear Data Library on BFS and ZPPR Analysis Results

Shono, Akira; Mantourov, G.

Nuclear Science and Engineering, 144(3), p.211 - 218, 2003/07

https://doi.org/10.13182/NSE03-A2354

None

Evaluation of nuclear constant effects relating to application of unified cross-section set ADJ2000

; Chiba, Go; Numata, Kazuyuki*

JNC TN9400 2002-062, 39 Pages, 2002/11

JNC-TN9400-2002-062.pdf:1.07MB

The unfied cross-section set ADJ2000, which was produced by adjusting 70-energy-grouped JFS-3-J3.2 (JENDL-3.2-based group constant set) using its analysis accuracy on various critical experiments, had been used in JFY2001 for core design studies in the feasibility study on commercialized fast reactor cycle systems. The ADJ2000 was published in June 2001, having 70 energy group structure as well. Calculated values from ADJ2000 would contain errors ("nuclear constant effects") which come from differences of weighting functions (neutron spectra) between that used for processing JFS-3-J3.2 and that of studied cores. In order to obtain more reliable results, the nuclear constant effects have to be corrected. Accordingly. nuclear constant effects on various fast reactor core concepts relating to application of unified cross-section set ADJ2000 have been evaluated for several nuclear parameters including criticality burnup reactivity loss, coolant void reactivity (depressurization reactivity in the case of gas-cooled reactor cores), Doppler reactivity and breeding ratio. Studied cores are listed below. (1)Sodium-cooled MOX-fuelled cores (large-size and middle-size) (2)Sodium-cooled metal-fuelled core (middle-size) (3)Lead-bismuth-cooled nitride-fuelled core (middle-size) (4)Carbon-dioxide-cooled MOX-fuelled core (large-size) (5)Helium-cooled coated-particle-fuelled core (large-size) (6)Helium-cooled enclosed-pin-fuelled core (large-size, 2 designs)

Analyses on the BFS critical experiments; an analysis on the BFS-62-3A and 62-4 cores

Hazama, Taira; ; Iwai, Takehiko*; Sato, Wakaei*

JNC TN9400 2002-036, 113 Pages, 2002/06

JNC-TN9400-2002-036.pdf:4.44MB

In order to support the Russian excess weapons plutonium disposition program, the intemational collaboration has started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering(IPPE). In the frame of the collaboration, analyses have been carried out for a series of critical experiments that simulate BN-600 (Russian commercial fast reactor). This report summarizes analysis results of the critical expeliments on BFS-62-3A and BFS-62-4 cores. BFS-62-3A core models BN-600 hybrid core in which the present BN-600 core is modified so as to partially load MOX fuel assemblies and replace the blanket region with stainless steel. BFS-62-4 core has the same layout as BFS-62-3A core except that the blanket region is not replaced. The analyses were performed with JNC standard method developed in the analysis of JUPITER experiment. The results show a good agreement with experimental values for the criticality and the reaction rate ratio. For the control rod worth and the reaction rate distribution, the results for BFS-62-4 core are also reasonable. However, for BFS-62-3A, analysis results overestimate the reaction rate in the stainless steel region by 20% and underestimate reactivity worth for one of the control rods by 10%. For the sodium void reactivity, underestimation of more than 20% were observed, but the disagreement were successfully solved by adopting a newly developed nuclear constant set with a fine group structure. In addition, analysis accuracies were compared among a series of analyses and it was confirmed that the introduction of MOX fuel assemblies does not affect the accuracy. The final goal of the work is to reflect the analysis results for designing BN-600 hybrid core. Then similarity was investigated between BFS-62-3A core and BN-600 hybrid core. A good similarity was found in the neutron spectrum, the fission reaction ratio, the fission reaction distribution, and the control rod worth. However, ...

Experimental Analysis Results on BN-600 Mock-up Core Characteristics at the BFS-2 Critical Facility

Shono, Akira; Sugino, Kazuteru; Hazama, Taira; Ishikawa, Makoto

7C-04, 0 Pages, 2002/00

None

None

Okawachi, Yasushi; ; Koshizuka, Seiichi*

JNC TY9400 2001-017, 117 Pages, 2001/05

JNC-TY9400-2001-017.pdf:3.3MB

no abstracts in English

Decay heat-measurement of minor actinides at YAYOI; Measurement results of gamma-ray decay heat

Okawachi, Yasushi;

JNC TN9400 2001-001, 100 Pages, 2000/08

JNC-TN9400-2001-001.pdf:2.83MB

Gamma-ray decay heat released from fission products has been measured for fast neutron fissions of U-235 and Np-237 using the radiation spectrometry method. The samples were irradiated at fast neutron source reactor "YAYOI" of the University of Tokyo. Gamma-ray energy spectra were measured using a NaI(TI) scintillation detector. And, the number of fission was evaluated from measured gamma spectra by Ge detector. For the measured gamma-ray, the background count was subtracted from the pulse height distribution of 1024 channels measured. The results were grouped by 340 channels to match the response matrix of the detector. This distribution was converted to energy spectra using the FERDO code and the response matrix of the detector. Normalized decay heat by the number of fission was obtained by integration of the energy spectra for each time step. The finite irradiation decay heat that is directly obtained by experiments can not be compared with experimental results and calculational results obtained under various irradiation conditions. So, the finite irradiation decay heat was converted to the fission burst decay heat. These results were compared with summation calculations using JNDC-V2 dacy data file. The present results on U-235 were compared with other experimental data using the same method. The present results agreed with other experimental data using the same method within 10%, suggesting the repeatability of experimental method. The present results on Np-237 were compared with the results of summation calculations using JNDC-V2 decay data file. As the result, the present results agreed with summation calculations within 8%. Probelms to be solved for the future are estimation of the experimental error, re-evaluation of the number of fission using updated nuclear data. To improve accuracy of decay heat data in shorter cooling time range, less irradiation experiment will be useful. Furthermore, to improve accuracy of decay heat data in longer cooloing ...

Experimental analyses results on the BFS 58-1-I1 critical assemblies

; Sato, Wakaei*; Iwai, Takehiko*

JNC TN9400 2000-096, 113 Pages, 2000/06

JNC-TN9400-2000-096.pdf:3.1MB

This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0 fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0 zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...