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Doda, Norihiro; Kato, Shinya; Iida, Masaki*; Yokoyama, Kenji; Tanaka, Masaaki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
In the conventional core design in sodium-cooled fast reactors (SFRs), negative reactivity feedback due to core deformation was neglected because of large uncertainty in analytical evaluation. To optimize core design, it is necessary to develop an analytical evaluation method and eliminate excessive conservativeness. An evaluation method for core deformation reactivity has been developed by coupling analysis of neutronics, thermal-hydraulics, and structural mechanics. For the verification study of the neutronics calculation method, the reactivity was calculated for the deformed core geometry in which the fuel assembly (FA) moved horizontally in the radial direction for each row from the core center. Compared to reference values derived from Monte Carlo calculations, the calculated reactivity due to FA displacement agreed well in the core region and was overestimated in the reflector region. The modification challenges in development of the core deformation model were identified.
Okonogi, Kazunari*; Nakamura, Takehiko; Yoshinaga, Makio; *
JAERI-Data/Code 99-018, 112 Pages, 1999/03
JAERI-Data-Code-99-018.pdf:4.48MB
no abstracts in English
Mizuno, Masahiro*; Uto, Nariaki
JNC TN9400 98-007, 147 Pages, 1998/11
A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under steady state and various transient operation modes. Heavy water is selected as a coolant material for heat removal of the SERAPH driver core during the experiments. Control rods are needed to conduct the experiments, and a control rod with heavy water follower is considered as one of the promising ideas and is now under investigation. In this idea, care must be taken to avoid production of local power peaks which are caused by neutron moderation in the follower and may appear in the vicinity of the boundary between the control rod and its neighboring fuel subassembly, since deuterium has an excellently high moderation power. Therefore, preparation of some methods of evaluating power density distribution in detail is required for control rod design. This report describes preparation of a set of neutronic calculation methods to evaluate intra-subassembly power density distribution including local power peaks around a control rod. A two-dimensional S
transport calculation code TWOTRAN-II is selected as a tool for evaluating neutron transport phenomena near the control rod with no cares for statistical influence. A two-dimensional rectangular super-cell model, which is a physical model composed of a control rod and its surrounding fifteen fuel sub-assemblies, and a method to construct the super-cell model based on thirteen unit cells are created, considering neutron mean free path near a control rod. Two processing tools are newly developed to generate a material region map and mesh boundaries for an efficient super-cell construction procedure and to obtain pin-wise power densities based on calculated mesh-wise neutron flux data. The results in this report are expected to be ...
Takeda, Toshikazu*; *; Kitada, Takanori*; *
PNC TJ9605 97-001, 100 Pages, 1997/03
This report is composed of the following two parts and appendix. (I)Improvement of the Method for Evaluating Reactivity Based on Monte Carlo Perturbation Theory (II)Improvement of Nodal Transport Method for 3-D Hexagonal Geometry (Appendix) Effective Cross Section of 
U Samples for Analyzing Doppler Reactivity in Fast Reactors Part I. Improvement of the Method for Evaluating Reactivity Bascd on Monte Carlo Perturbation Theory. Theoretical formulation in Monte Carlo perturbation method had been checked, and then introduced into a calculation code. The increase of CPU time is about 10 to 20 % compared to that if normal Monte Carlo code, in the cases of same number of history. This Monte Carlo perturbation method found to be effective, because results are almost reasonable and deviations of the results are especially small, by using the Monte Carlo perturbation code. However, there are somc cases that the results of the change of eigenvalues becomes positive or negative by changing the estimator, and there is no reasonable difference in the results between the conventional method, which does not consider the change of neutron source distribution caused by a perturbation, and the new method, which consider that change. Thus it is still necessary to check the Monte Carlo pcrturbation code. Part II. Improvement of Nodal Transport Method for 3-D Hexagonal Geometry It is certain that we can accurately evaluate hexagonal geometry FBR core by nodal transport calculation code for hexagonal-Z geometry named 'NSHEX'. However it is also found that in very heterogeneous core the results is not good enough. Because the treatment of the transverse leakage to the radial direction, which is use for evaluating intra-nodal flux distribution, is not so accurate. For the treatment of the leakage distribution, it is necessary to estimate the nodal vertex flux. In conventional method, the vertex flux estimated by the surrounding node surfacc flux around that vertex. On the contrary,
Iijima, Susumu; Okajima, Shigeaki; Nakakawa, Masayuki
Nihon Genshiryoku Gakkai-Shi, 39(1), p.25 - 27, 1997/00
no abstracts in English
; Nomura, Masahiro;
PNC TN9410 96-205, 42 Pages, 1996/07
At present, a high power CW(Continuous Wave) electron linac accelerator is under development in PNC as a part of transmutation study by accelerators. As one of applied uses, we study a hybrid reactor system combining the linac and a subcritical core with TRU fuel. This report shows nuclear and thermal calculation results to obtain characteristics of the system. In the system, an electron beam from the linac is injected into a target located in the center of the system. In the first place, the injected electrons generate photons of 
ray via Bremsstrahlung production. After this, these photons produce neutrons by (, n) reactions. As a result, an extinguished quantity of TRU fuel is about 0.1kg when the electron beam of incident energy 100 MeV is injected into the target over 1 year. It is proven that the system might be established from the point of thermal engineering. The hybrid reactor system can also use a proton accelerator instead of the electron one. Nuclear and thermal calculation will be performed on the system using the proton accelerator. Comparison of the extinguished quantity of TRU fuel will be performed between calculation results of both systems.
Okumura, Keisuke; Kugo, Teruhiko
RIST News, (20), p.20 - 25, 1995/00
no abstracts in English
*; *; Kosako, Kazuaki*; Seki, Yasushi
JAERI-M 91-073, 101 Pages, 1991/05
no abstracts in English
*; Kosako, Kazuaki*; Seki, Yasushi; *
JAERI-M 91-072, 103 Pages, 1991/05
no abstracts in English
; Nakakawa, Masayuki; Mori, Takamasa; Kugo, Teruhiko; *
JAERI-M 90-177, 96 Pages, 1990/10
no abstracts in English
; Okumura, Keisuke; Takano, Hideki;
JAERI-M 90-109, 49 Pages, 1990/07
no abstracts in English
Takeda, Toshikazu*; Unesaki, Hironobu*; Kurisaka, Kenichi*; Sakuma, Hiroomi*; Shimoda, Masayuki*; Ito, Noboru*; Kugo, Teruhiko*; Aoki, Shigeaki*; Uto, Nariaki*; Tanaka, Motonari*
PNC TJ2605 88-001, 230 Pages, 1988/03
no abstracts in English
; Tanaka, Kichizo;
JAERI-M 6150, 18 Pages, 1975/06
no abstracts in English
; Katsuragi, Satoru
JAERI-M 5381, 60 Pages, 1973/08
no abstracts in English
Nagaya, Yasunobu
no journal, ,
This is a textbook for a lecture entitled with "Neutronics calculation with Python, 2; Stochastic method", which will be presented at the summer seminar of reactor physics organized by Atomic Energy Society of Japan, Reactor physics division. Fundamentals of Monte Carlo methods including random number generation and sampling method are illustrated with sample codes of the Python language. Monte Carlo algorithms are also described for a fixed-source problem, 1-group/2-group eigenvalue problems for simple spherical geometry.
Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
JAEA has developed a multi-level plant simulation system to enhance a safety assessment technology for sodium cooled fast reactor (SFR). The system, in which high-efficient plant dynamics analyses with a one-dimensional system code and high-accurate local phenomena analyses with multi-dimensional codes are coupled, may apply to the various design options of SFR. We have developed two coupling methods with plant dynamics analysis code for fuel assembly thermal-hydraulics analysis code and for neutronics analysis code, respectively. Using the coupling methods, we performed analyses on the thermal-hydraulics coupling problem between whole core and a fuel assembly and on the nuclear-thermal coupling problem under unprotected conditions. The coupled analysis results were compared with the 1D code results of the same problems. The results showed that each coupling method was validated.