Validation of MOSRA-SRAC for burnup of a BWR fuel assembly

Kojima, Kensuke

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3283 - 3292, 2016/05

The MOSRA system has been developing to improve the applicability of the neutronic characteristic analyses. The cell calculation module MOSRA-SRAC is a core module of MOSRA, and applicability tests for realistic problems are required. As a test, MOSRA-SRAC is validated by comparison with measured values. As the measurement, the post irradiation examination SFCOMPO 99-5 is chosen. In the examination, the compositions of major heavy metal and fission product nuclides in a UO-GdO fuel rod pulled from the 88 BWR fuel assembly used in TEPCO's Fukushima-Daini-2 were measured. The result shows good agreement between calculated and measured value. For uranium and plutonium nuclides, calculated values agree within 5% except for Pu. Pu composition is overestimated by 30%, and the overestimation is caused by the unclearness of the void faction history of the fuel rod. For fission products, calculated values agree within approximately 10%.

Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki*; Torii, Kazutaka*

JAEA-Research 2015-019, 90 Pages, 2016/01

JAEA-Research-2015-019.pdf:1.95MB

At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For the purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core.

Benchmark calculation with MOSRA-SRAC for burnup of a BWR fuel assembly

Kojima, Kensuke; Okumura, Keisuke

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 9 Pages, 2014/09

The MOSRA system has been developing to improve the applicability of the neutronic characteristic analyses. The cell calculation module MOSRA-SRAC is a core module of MOSRA, and applicability tests for realistic problems are required. As a test, we joined the benchmark "Burnup Credit Criticality Benchmark Phase IIIC." The benchmark requested the neutronic characteristics for a BWR fuel assembly with gadolinium, which had been used in the TEPCO's Fukushima Daiichi Nuclear Power Station. Because of a restriction of MOSRA-SRAC, the geometry was partially homogenized. To verify the module's applicability including the homogenization effects, the multiplication factor and the nuclide compositions were compared with the well-validated code MVP-BURN. As the results, the applicability of MSORA-SRAC for the assembly was verified. Additionally, it was also shown that the homogenization effects were smaller than the difference due to the calculation methods.

Main generation pathways of activation products for in-core structure materials (Joint research)

Yamamoto, Kento; Okumura, Keisuke; Kojima, Kensuke; Okamoto, Tsutomu

JAEA-Research 2013-038, 88 Pages, 2014/02

JAEA-Research-2013-038.pdf:3.15MB

Accurate information on main generation pathways of activation products is important to improve the accuracy of predicting the concentrations of activation products. Using ORIGEN2 and ORLIBJ40, which is a set of the cross section libraries based on JENDL-4.0, the initial compositions and the cross sections which influence on the concentrations of activation products were clarified by executing the sensitivity analyses on them. Activations of cladding tubes, end-plugs and spacers of fuel assemblies and channel boxes, which were composed of Zircaloy, SUS304, and Inconel-718, were analyzed. The main generation pathways of some significant activation products were summarized from the results of sensitivity analyses.

Nuclear data for severe accident analysis and decommissioning of nuclear power plant

Okumura, Keisuke; Kojima, Kensuke; Okamoto, Tsutomu; Hagura, Hiroyuki; Suyama, Kenya

JAEA-Conf 2013-002, p.15 - 20, 2013/10

https://doi.org/10.11484/jaea-conf-2013-002

Three-dimensional nuclide inventory and decay heat analysis were performed for the Fukushima Dai-ichi Power Plants (1F1, 1F2, 1F3) by using MOSRA system with JENDL-4.0 library. In the analysis, nuclide inventory for approximately 1400 nuclides were estimated in consideration of radial and axial burn-up and void distributions. Total decay heat and its distribution in each plant were estimated by the sum of all nuclide contributions. The obtained decay heat was compared with the results of simple evaluation formulas used in severe accident analyses. The results of the simple evaluation formulas agree with our results within 20%. For future decommissioning of commercial nuclear power plants, new activation cross-sections library for ORIGEN-S is also under development in the cooperative study program between JAEA and JAPCO. The present status and future plan are shown from view points of nuclear data and method.

Production of the ORIGEN2 library based on JENDL-4.0 for high temperature engineering test reactor

Kojima, Kensuke; Okumura, Keisuke; Okamoto, Tsutomu; Goto, Minoru

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 7 Pages, 2012/10

A set of the ORIGEN2 library for High Temperature engineering Test Reactor (HTTR) was newly produced in order to improve prediction accuracy of burn-up characteristics such as spent fuel composition, radioactivity and decay heat. In the library, cross sections and decay data are adopted from JENDL-4.0 and from ENSDF. In the production of effective cross sections and neutron spectrum, MVP-BURN based on the continuous-energy Monte Carlo method and a statistical geometry model is applied to the HTTR fuel with many coated fuel particles. In this way, the double heterogeneous effect of the HTTR fuel can be accurately taken into account. By using the neutron spectrum obtained in the MVP-BURN calculation, infinite dilution cross sections from JENDL-4.0 are condensed to one-group cross sections. The burn-up calculation results of ORIGEN2 with the produced library and those of MVP-BURN with detail modeling and much calculation cost show good agreement for burn-up changes of fuel composition.

Development of calculation system for decontamination effect, CDE

Satoh, Daiki; Kojima, Kensuke; Oizumi, Akito; Matsuda, Norihiro; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

JAEA-Research 2012-020, 97 Pages, 2012/08

JAEA-Research-2012-020.pdf:7.32MB

A computer software, named CDE (Calculation system for Decontamination Effect), has been developed to support planning the decontamination. CDE calculates the dose rates before the decontamination by using a database of dose contributions by radioactive cesium. The decontamination factor is utilized in the prediction of the dose rates after the decontamination, and dose rate reduction factor is evaluated to express the decontamination effect. The results are visualized on the image of a target zone with color map. In this paper, the overview of the software and the dose calculation method are reported. The comparison with the calculation results by a three-dimensional radiation transport code PHITS is also presented. In addition, the source code of the dose calculation program and user's manual of CDE are attached as appendices.

New ORIGEN2 libraries based on JENDL-4.0 and their validation for long-lived fission products by post irradiation examination analyses of LWR spent fuels

Kojima, Kensuke; Okumura, Keisuke; Asai, Shiho; Hanzawa, Yukiko; Okamoto, Tsutomu; Toshimitsu, Masaaki; Inagawa, Jun; Kimura, Takaumi; Kaneko, Satoru*; Suzuki, Kensuke*

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Accurate inventory estimation of long-lived fission products (LLFPs) in LWR spent fuels is important for the quality management and for long-term safety assessment of high-level radioactive vitrified wastes. In Japan, ORIGEN2 has been widely used to estimate the fuel compositions. However, equipped library data in the original ORIGEN2 are old and are not validated enough for LLFPs, such as Se, Tc, Sn and Cs, because available post irradiation examination (PIE) data are limited for these nuclides, which have difficulties in radiochemical analyses. For more accurate the estimation, new ORIGEN2 libraries are developed from the latest nuclear data library JENDL-4.0 for cross sections and fission yields, and from other libraries for half-lives, and so on. The new libraries are validated by PIE analyses of the sample fuels irradiated in Cooper, Calvert-Cliffs-1, H. B. Robinson-2, and Ohi-1. As a result, it was found that the new library gives good results for the estimation.

Application of bias factor method with use of exponentiated experimental value to prediction uncertainty reduction in coolant void reactivity of breeding light water reactor

Kugo, Teruhiko; Kojima, Kensuke; Ando, Masaki; Mori, Takamasa; Takeda, Toshikazu*

Journal of Power and Energy Systems (Internet), 2(1), p.73 - 82, 2008/00

https://doi.org/10.1299/jpes.2.73

We have applied the bias factor method to coolant void reactivity of a breeding light water reactor with use of FCA-XXII-1 experiment with introducing a concept of exponentiated experimental value into the bias factor method in order to overcome a problem caused by the conventional bias factor method in which the prediction uncertainty increases in the case that the experimental core has the opposite reactivity worth and the consequent opposite sensitivity coefficients to the real core. In the present study, we have formulated the prediction uncertainty reduction by the use of the bias factor method extended by the concept of the exponentiated experimental value. From the numerical results, it is verified that the concept of exponentiated experimental value can improve the prediction accuracy compared with the original uncertainty in the design calculation value while the conventional bias factor method cannot improve the prediction accuracy. It is concluded that the introduction of exponentiated experimental value can effectively utilize experimental data and extend applicability of the bias factor method.

Preliminary evaluation of reduction of prediction error in breeding light water reactor core performance

Kugo, Teruhiko; Kojima, Kensuke; Ando, Masaki; Okajima, Shigeaki; Mori, Takamasa; Takeda, Toshikazu*; Kitada, Takanori*; Matsuoka, Shogo*

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 10 Pages, 2005/05

We have preliminarily evaluated the reduction of prediction errors of the core characteristics of the breeding light water reactor core based on the bias factor method by utilizing the FCA critical experiments carried out for MOX fueled tight lattice light water reactor cores. The prediction uncertainty of k is reduced from 0.62% to 0.39% by utilizing the FCA-XV-2 (65V) result. As for the reaction rate ratio of U capture and Pu fission, it is found that the FCA XXII-1 (95V) and XV (95V) results are suitable for the upper core and the upper blanket of the real core and the FCA XXII-1 (65V) and XV-2 (65V) results are suitable for the lower core and the internal blanket.