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JAEA Reports

Applicability of statistical geometry model to light water moderating systems

Mori, Takamasa; Kojima, Kensuke*; Suyama, Kenya

JAEA-Research 2018-010, 57 Pages, 2019/02

JAEA-Research-2018-010.pdf:6.25MB

In order to estimate applicability of the statistical geometry model (STGM) of MVP/GMVP, a parametric study in infinite geometry and criticality safety analyses for direct disposal of spent fuel in simple finite geometry have been carried out by using the MVP Monte Carlo code. It has been found that calculations with STGM for larger fuel spheres give larger thermal utilization factors and larger infinite multiplication factors compared with explicit random models in the range of fuel sphere packing fraction between 6.5 % and 63.3 %. Substantial differences are not observed between the results with two nearest neighbor distributions (NNDs); that given by the MCRDF code and the analytical expression based on a statistically uniform distribution. It is inferred that the overestimation by STGM is caused by the facts that STGM cannot take account of the surroundings of each neutron, whether a fuel sphere rich region or a water moderator rich one, because STGM always uses an NND averaged over such surroundings and that STGM, therefore, cannot take the effect of consecutive scatterings in the water moderator into account.

Journal Articles

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$$_{2}$$-Gd$$_{2}$$O$$_{3}$$ fuel rod pulled from the 8$$times$$8 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 $$^{238}$$Pu. $$^{238}$$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%.

Journal Articles

Development of multi-group neutron activation cross-section library for decommissioning of nuclear facilities

Okumura, Keisuke; Kojima, Kensuke; Tanaka, Kenichi*

JAEA-Conf 2015-003, p.43 - 47, 2016/03

In the safety assessment concerning disposal of radioactive wastes generated in the decommissioning of nuclear facilities, it is necessary to evaluate the radionuclide inventory produced by the activation of structured materials. For this purpose, we have to pay much attention to the activation of many impurities irradiated in various neutron spectra depending on their positions and materials. Therefore, accurate activation cross-section data are necessary for many nuclides and reactions. A new multi-group neutron activation cross-section library (MAXS) was developed based on the recent nuclear data JENDL-4.0 and JEFF-3.0/A to apply it to the activation calculations for the decommissioning of nuclear facilities. The library contains cross-sections and isomeric ratios for many reactions such as (n,$$gamma$$), (n,f), (n,2n), (n,3n), (n,p), (n,$$alpha$$), (n,d), (n,t), (n,n$$alpha$$), (n,np), and so on, for 779 nuclides, in the 199-energy group structure of VITAMIN-B6.

JAEA Reports

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.

Journal Articles

Sensitivity analyses of initial compositions and cross sections for activation products of in-core structure materials

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

Nuclear Back-end and Transmutation Technology for Waste Disposal, p.233 - 249, 2015/00

To improve the prediction accuracy of concentrations of activation products (APs) in the field of nuclear back-end, it is important to investigate the elements and the nuclear reactions leading to generation of APs. To clarify quantitatively the source elements and the nuclear reactions dominating generation of APs, sensitivity analyses of initial compositions and cross sections were conducted using ORIGEN2.2 code and ORLIBJ40, which is a set of the 1-group cross section libraries based on JENDL-4.0. Activations of cladding tubes, end-plugs and spacers of fuel assemblies and channel boxes in BWR, whose materials are zirconium alloy, stainless steel, and nickel-chromium-based alloy, were analyzed. The results clarified quantitatively the source elements and the nuclear reactions dominating generation of APs. It was remarkable that the dominant generation pathways were clarified even for the nuclides generated through complicated pathways. In conclusion, the results of sensitivity analyses could be utilized to select the objective of elements for measurements of impurities in the materials and of nuclear data for improvement of accuracy.

Journal Articles

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.

Journal Articles

Development of a calculation system for the estimation of decontamination effect

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

Journal of Nuclear Science and Technology, 51(5), p.656 - 670, 2014/05

 Times Cited Count:3 Percentile:61.51(Nuclear Science & Technology)

A calculation system for the estimation of decontamination effect (CDE) has been developed to support planning a rational and effective decontamination. The method calculates the dose-rate distribution before and after decontamination, according to the distribution of radioactivity and the decontamination factor (DF), and uses a dose rate reduction factor (DRRF) to estimate the decontamination effect. The results that were calculated by using the CDE were compared with the results of measurements as well as with the results of calculations that were performed using a Monte Carlo particle transport code PHITS. It was found that the CDE successfully reproduced the measured as well as the calculated dose-rate distributions, requiring less than several seconds of calculation time.

JAEA Reports

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.

Journal Articles

Calculation system for the estimation of decontamination effect

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

Transactions of the American Nuclear Society, 109(1), p.1261 - 1263, 2013/11

A computer software, named CDE (Calculation system for Decontamination Effect), has been developed to support planning the decontamination. CDE is programed with VBA (Visual Basic for Applications), and runs on Microsoft Excel with a user friendly graphical interface. It calculates dose rate distributions in a target area before and after the decontamination from a radioactivity distribution and DF (Decontamination Factor), which is a ratio of original radioactivity to remaining one after the decontamination. DRRF (Dose Rate Reduction Factor) is also derived to express the decontamination effect. All the calculation results are visualized on an image of the target area with color map. Owing to its quick calculation speed, CDE is able to investigate the decontamination effect in various cases for a short period. This is very useful to establish a rational decontamination plan before an action.

Journal Articles

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

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.

JAEA Reports

A Set of ORIGEN2 cross section libraries based on JENDL-4.0; ORLIBJ40

Okumura, Keisuke; Sugino, Kazuteru; Kojima, Kensuke; Jin, Tomoyuki*; Okamoto, Tsutomu; Katakura, Junichi*

JAEA-Data/Code 2012-032, 148 Pages, 2013/03

JAEA-Data-Code-2012-032.pdf:6.99MB

A set of cross section libraries for the isotope generation and depletion calculation code ORIGEN2 was produced by using recent nuclear data JENDL-4.0. In this new library (ORLIBJ40), neutron-induced cross sections, fission product yields, isomeric ratios and half-lives were updated. ORLIBJ40 includes 24 libraries for typical UO$$_{2}$$ or MOX fuels of PWR and BWR. In addition, it includes 36 libraries for various fast reactor fuels. ORLIBJ40 was applied to the post irradiation examination analyses of LWR nuclear spent fuels. As a result, it was confirmed that improvements were achieved especially for inventory and radioactivity estimations of minor actinides (Am and Cm isotopes) and fission products sensitive to cross sections (Eu and Sm isotopes) and for long-lived fission products ($$^{79}$$Se, etc.), compared with other existing ORIGEN2 libraries.

Journal Articles

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.

Journal Articles

Decontamination planning based on computer simulation code CDE

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

RIST News, (53), p.12 - 23, 2012/09

Decontamination planning based on a computer simulation code CDE is discussed in this paper. Large amount of radionuclides had been discharged to environment in the accident of the Tokyo Electronic Power Corporation Fukushima Dai-ichi Nuclear Power Plant. CDE has been developed to support planning the decontamination. From the present study, it is validated that the computer simulation is very useful to predict the effect of the scenario before actions, and to plan the decontamination.

JAEA Reports

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.

Journal Articles

Development of the ORIGEN2 library for light water reactors based on JENDL-4.0

Okumura, Keisuke; Kojima, Kensuke; Okamoto, Tsutomu

JAEA-Conf 2012-001, p.89 - 94, 2012/07

New ORIGEN2 libraries (ORLIBJ40) for light water reactors were developed based on JENDL-4.0. It was validated by post irradiation examination analyses of nuclear spent fuels. As a result, it was found that drastic improvements were achieved especially for the inventory estimations of minor actinides (Np, Am and Cm isotopes) and fission products sensitive to cross sections (Eu and Sm isotopes) compared with old ORIGEN2 libraries. Furthermore, radioactivity of long-lived fission products, such as $$^{79}$$Se and $$^{135}$$Cs were also improved because of implementations of new half-life data. These nuclides are very important for long-term safety assessment of a geological disposal of the vitrified wastes.

Journal Articles

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 $$^{79}$$Se, $$^{99}$$Tc, $$^{126}$$Sn and $$^{135}$$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.

Journal Articles

Prediction accuracy improvement of neutronic characteristics of a breeding light water reactor core by extended bias factor methods with use of FCA-XXII-1 critical experiments

Kugo, Teruhiko; Ando, Masaki; Kojima, Kensuke; Fukushima, Masahiro; Mori, Takamasa; Nakano, Yoshihiro; Okajima, Shigeaki; Kitada, Takanori*; Takeda, Toshikazu*

Journal of Nuclear Science and Technology, 45(4), p.288 - 303, 2008/04

 Times Cited Count:6 Percentile:51.42(Nuclear Science & Technology)

The effectiveness of the extended bias factor methods, the LC and PE methods, is numerically investigated by applying them to a breeding light water reactor core as a target core with use of FCA-XXII-1 critical experiments. The present study numerically verifies the features of the extended bias factor methods. Both the methods can improve the prediction accuracy the most by using all the experiments. The PE method always improves the prediction accuracy with any combination of experiments. The PE method is always superior to the LC method for improvement of the prediction accuracy. From the present study, the followings are found. The experiments on multiplication factor are more applicable to a reaction rate ratio of $$^{238}$$U capture to $$^{239}$$Pu fission (C28/F49) of the target core than the experiments on C28/F49. Combinations of the experiments on multiplication factor is more effective to a void reactivity of the target core than those of the experiments on void reactivity though those on void reactivity are superior to those on multiplication factors in the case of using a single experiment. From these results, we conclude that the experiments on multiplication factor are more effective than the other experiments for all the neutronic characteristics of the target core. From these results, it is concluded that the PE method is promising to complement full mockup experiments for various future nuclear systems by using a number of existing and future benchmark experiments.

Journal Articles

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

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.

Journal Articles

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

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

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04

Utilizing the critical experiments for MOX fueled tight lattice LWR cores at FCA XXII-1 cores, we have evaluated prediction uncertainty reduction in coolant void reactivity worth of a breeding LWR core based on the bias factor method. In the present study, to extend the applicability of the bias factor method, we have introduced an exponentiated experimental value as a virtual experimental value and formulated the prediction uncertainty reduction with the bias factor method extended by the concept. From the numerical evaluation, it has been shown that the prediction uncertainty due to cross section errors has been reduced by the use of the concept of the virtual experimental value. It is concluded that the introduction of virtual experimental value can effectively utilize experimental data and extend applicability of the bias factor method.

Journal Articles

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$$_{eff}$$ is reduced from 0.62% to 0.39% by utilizing the FCA-XV-2 (65V) result. As for the reaction rate ratio of $$^{238}$$U capture and $$^{239}$$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.

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