Development of a unified cross-section set ADJ2010 based on adjustment technique for fast reactor core design

Sugino, Kazuteru; Ishikawa, Makoto; Yokoyama, Kenji; Nagaya, Yasunobu; Chiba, Go; Hazama, Taira; Kugo, Teruhiko; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*

Journal of the Korean Physical Society, 59(2), p.1357 - 1360, 2011/08

https://doi.org/10.3938/jkps.59.1357

In order to improve the prediction accuracy of core performances in the fast reactor core design study, the unified cross-section set has been developed in Japan. The unified cross-section set, which combines a wide range of integral experimental information with differential nuclear data, is produced by using the cross-section adjustment technique based on the Bayesian parameter-estimation theory. A new set ADJ2010 is currently under development. The present paper describes the results of the cross-section adjustment for ADJ2010 which is based on the JENDL-4.0 data. The evaluation of the core design accuracy for a commercial power fast reactor core is also discussed. ADJ2010 will be released soon and will be expected to be utilized for core design of future fast reactors.

Simultaneous recovery of all actinides from spent nuclear fuel by carbamoylmethylphosphine oxide in fluorinated diluents

Ozawa, Masaki; Iwai, Takehiko*; Babain, V.*; Shadrin, A.*

Proceedings of International Solvent Extraction Conference "Solvent Extraction-Fundamentals to Industrial Applications" (ISEC 2008), p.623 - 628, 2008/09

Bifunctional organophosphorus extractants dissolved in polar fluorinated diluents were studied aiming at directly recovering all f-elements from the dissolver solution of spent nuclear fuel. Distribution ratios of U, Np and Pu were sufficiently high for 0.40.8MCMPO in this solvent system, and combination of salt-free, methylamine carbonate (MAC), etc, were evaluated to obtain fractional stripping of f-elements. Static multi-stage extraction using artificial FBR dissolver solution supported the process feasibility.

MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR Nuclear core design

Sugino, Kazuteru; Iwai, Takehiko*

Proceedings of American Nuclear Society Topical Meeting on Physics of Reactors (PHYSOR 2006) (CD-ROM), 10 Pages, 2006/09

MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores in Japan. Subsequently, feasibility of MONJU integral data was evaluated by cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to remarkably improve the prediction accuracy of neutronic parameters for MONJU.

Results of Nuclear Design Accuracy Evaluation on BN-600 Hybrid Core

Shono, Akira; Sato, Wakaei*; Hazama, Taira; Iwai, Takehiko*; Ishikawa, Makoto

JNC TN9400 2003-074, 401 Pages, 2003/08

JNC-TN9400-2003-074.pdf:48.95MB

Nuclear design accuracy on the BN-600 hybrid core has been evaluated using the JNC's nuclear analysis system for FBR cores, by utilizing the critical experiment analysis results on BFS-62 configuration that had been obtained under JNC's efforts for Russian surplus weapons plutonium disposition. In the BN-600 hybrid core design, a part of the current UO2 fuel region is replaced by MOX fue1, and the Peripheral blanket region by stainless steel reflectors, respectively. These changes were simulated in a series of critical experiment configurations (BFS-62-1 to 4). Based on the analysis results on both BFS-62 configurations and other fast reactor cores, nuclear design accuracy on the BN-600 hybrid core has been evaluated by applying both the group constant adjustment method and the bias method. Evaluated nuclear parameters include, the criticality, fission rate distribution, sodium void reactivity, control rod worth, burn-up reactivity loss, etc. It is concluded, by applying the group constant adjustment method, that the evaluated accuracy (uncertainty) of most of the nuclear parameters can be decreased to less than half of those based on the basic nuclear constant without reflecting any experimental data. The improvement was mainly achieved by reducing the covariance of the iron elastjc cross section. This significant effect results from the feature of the BN-600 hybrid core, which has relatively larger power density, adopts U235 as the main fissile nucljde, and has the stainless steel reflector surrounding the fuel region. In addition, good consistency of analysis results between the BFS and other fast reactor cores is confirmed. Information obtained by BFS-62 experiment show significant contribution to the accuracy improvement. It is also found that the bias method shows less significant effects on the accuracy improvement than the group constant adjustment method. Furthermore, the bias method may degrade the accuracy for certain nuclear parameters that have large e

Analyses on the BFS critical experiments; an analysis on the BFS-62-1 and 62-2 cores

Sugino, Kazuteru; ; Iwai, Takehiko*; Numata, Kazuyuki*

JNC TN9400 2002-008, 241 Pages, 2002/04

JNC-TN9400-2002-008.pdf:6.84MB

In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engneering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 and BFS-62-2 cores. The BFS-62-1 core models the present BN-600, and contains the enriched UO fuel surrounded by the UO blanket. The BFS-62-2 core has the same layout as the BFS-62-1 but the blanket region was replaced with stainless steel shield. For core parameter analyses, the 3-D Hexagonal-Z or XYZ geometry model was applied by not only diffusion calculation but also transport calculation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality, the reaction rate ratio and reaction rate distribution in BFS-62-1. In the reaction rate distribution of BFS-62-2 calculation without cross-section adjustment produced big radial dependency of calculation over experiment value (C/E value) in the core region and overestimation in the shield region. Cross-section adjustment technique procedure improved those estimation, however alternation of cross-section of Iron, which was dominant in above improvement, compared to the cross-section error, and further investigation was required. Concerning the control rod worth of BFS-62-1, radial dependency of the C/E value was observed whether cross-section adjustment technique was applied or not, therefore comparison with results of other BFS-62 ...

Analyse on the BFS critical experiments; An analysis on the BFS-62-1 assembly

Sugino, Kazuteru; Iwai, Takehiko*;

JNC TN9400 2000-098, 182 Pages, 2000/07

JNC-TN9400-2000-098.pdf:5.74MB

In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 assembly, which is the first core of the BFS-62 series. The core contains the enriched U0 fuel surrounded by the U0 blanket. The standard analytical method for fast reactors has been applied, which was used for the JUPITER and other experimental analyses. Due to the lack of the analytical data the 2D RZ core calculation was mainly used. The 3D XYZ core calculation was applied only for the preliminary evaluation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality and the reaction rate ratio. However, it was found that accurate evaluation of the reaction rate distribution was impossible without exact consideration of the arrangement of the two types of sodium (with and without hydrogen impurity), which can be accommodated by the 3D core analysis, thus it was essentia1. In addition, it was clarifie that there was a room for an improvement of the result on the reaction rate distribution in the blanket and shielding regions. The application of the 3D core calculation improved the result on the control rod worth because 3D core model can more exactly consider the shape of the control rod. Furthermore it was judged that the result of the analysis on the sodium void reactivity .....

MA transmutation in various fast reactor core concepts

Oki, Shigeo; Iwai, Takehiko*; Jin, Tomoyuki*

JNC TN9400 2000-080, 532 Pages, 2000/03

JNC-TN9400-2000-080.pdf:14.98MB

Transmutation Property of minor actinide nuclides (MA) was analyzed for fast reactor cores having different coolant and fuel types in order to obtain basic data for evaluating an ability of the efficient use of resource and reducing the effect on the environment. The investigated fast reactor cores were (a) sodium cooled and oxide fueled core, (b) lead cooled and nitride fueled core (BREST-300), (c) carbon dioxide gas cooled and oxide fueled core (ETGBR), (d)lead cooled and oxide fueled core, (e) sodium cooled and nitride fueled core (both He-bond and sodium-bond), and (f) sodium cooled and metallic fueled core. Followings were observed for the relation between MA transmutation property and the different types of coolant and fuel of fast reactor core. (1) For the MA transmutation rate, the relation "Oxide Metal Nitride" was found out for difference of fuel type. A main reason of the increment of MA transmutation rate is that the neutron flux level rises on nitride and metallic fueled cores in comparison with oxide core. (2) The relation "Lead Sodium and Carbon dioxide" can be seen for the MA transmutation rate in the difference of coolant, but it is not clear whether the cause is driven from the difference of coolant itself on the difference of core design. (3) The changes of MA transmutation property mentioned above are comparatively small.

Development of the ZPPR-9 core benchmark problem

Iwai, Takehiko*; Sugino, Kazuteru; Ishikawa, Makoto

PNC TN9410 98-079, 54 Pages, 1998/07

PNC-TN9410-98-079.pdf:1.37MB

In the Reactor Physics Research Section, O-arai Engineering Center, a Standard Data Base for Nuclear Design is under development to improve the accuracy of FBR nuclear design calculations. The Reactor Integral Test Working Group of the Sigma committee is compiling data from TCA (Tank Type Critical Assembly), STACY (Static Experimental Critical Facility) and FCA (Fast Critical Assembly), for a set of nuclear data benchmark problems. An FBR benchmark problem from the Working Group has been added to the Data Base: the ZPPR-9 core, simplest of the JUPITER-I series, recommended for its ease of use. The compiled nuclear parameters are Criticality; Reaction rate data(Reaction rate ratio & Reaction rate distribution); Control rod worth; Sodium void reactivity and Sample Doppler reactivity. The benchmark problem definition is an idealization of the experimental geometly; we used detailed analytical methods to prepare correction factors, so that users of the benchmark can compare their results with the experiment. Wide use of the benchmark problem is anticipated. However, the calculation of correction factors is affected by the mesh size, the number of energy groups and the definition of cross sections, so it is necessary to use detailed analytical methods to produce modified correction factors when one uses a calculational model in which mesh size etc. are different.

Development of a standard data base for FBR core nuclear design, VIII; Compilation of JUPITER analytical results

Ishikawa, Makoto; Sato, Wakaei*; Sugino, Kazuteru; Yokoyama, Kenji; Numata, Kazuyuki*; Iwai, Takehiko*

PNC TN9410 97-099, 512 Pages, 1997/11

PNC-TN9410-97-099.pdf:13.84MB

A Standard data base for LMFBR core nuclear design has been developed to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as demonstration or commercial FBRs. To develop the data base, extensive work has been prerformed to accumulate and evaluate many kinds of results from fast reactor physics experiments and their analyses. The present report summarizes the analytical results of the JUPITER experiments, using the most recent nuclear data library (JENDEL-3.2) and the lates analytical methods in a consistent manner. In the present work, a great emphasis was placed on guaranteeing the essential requirements for this kind of general data base, that is, "Accountability", "Traceability" and "Consistency". In other words, consistent strategies and analytical methods were applied to all calculations, including detialed corrections; the enormous analytical input data generated were all saved in the form of computer files, so that reanalysis of any experiment could be easily performed for verification or in response to future improvement in nuclear data or analytical methods. The main results of the present JUPITER analysis are as follows: (1) The C/E(calculation/experiment) values of criticality were slightly underestimated by -0.7-0.3%k. (2) The reaction rate ratio of C28/F49 was overestimated by +4+6% with the standard analytical method. However it was found to improve about 2% after the cell factors were revised using the Monte Carlo method. (3) The radial space-dependency of the reaction rate distribution and control rod worth almost disappeared in the homogeneous cores. (4) The previous overestimation of sodium void reactivity was greatly improved in the homogeneous cores.

Manual of standard procedures for analysis of radioactive effluents and gases discharged into environment

Iwai, Makoto; Akutsu, Ken*; *; *; Shimizu, Takehiko*; Hayashi, Naomi*; *; Maruo, Yoshihiro

PNC TN852 84-06, 379 Pages, 1984/03

PNC-TN852-84-06.pdf:11.08MB

This manual includes the standard procedures for analysis of radioactive materials and chemical pollutants in liquid and gaseous waste effluent discharged into the environment from the nuclear facilities of Tokai Works of PNC. The second edition, PNCT852 - 79 - 10 was published in 1979. Almost all analytical procedures have been modified and new technique has been used for these five years, so the third edition revised entirely was to be published this time. The principal consideration was placed on the revision of sample preparation and safety handling of the chemical agents in particular.