A Design study on a metal fuel fast reactor core for high efficiency minor actinide transmutation by loading silicon carbide composite material

Ohgama, Kazuya; Hara, Toshiharu*; Ota, Hirokazu*; Naganuma, Masayuki; Oki, Shigeo; Iizuka, Masatoshi*

Journal of Nuclear Science and Technology, 59(6), p.735 - 747, 2022/06

https://doi.org/10.1080/00223131.2021.2005702

Detail analysis for a control rod worth of the Gas Turbine High Temperature Reactor (GTHTR300)

Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, X.; Kunitomi, Kazuhiko

JAERI-Tech 2002-087, 83 Pages, 2002/11

JAERI-Tech-2002-087.pdf:3.47MB

no abstracts in English

Neutronic study of a very small reactor MR-1G core for exclusive use of heat supply

Odano, Naoteru; Ishida, Toshihisa; Ochiai, Masaaki

JAERI-Research 2001-044, 53 Pages, 2001/10

JAERI-Research-2001-044.pdf:2.35MB

A very small reactor, MR-1G with a thermal output of 1 MW, is the integral-pressurized type reactor to be used for heat supply to an office building in a city. Neutronic study has been carried out for design of the MR-1G, the core of which could achieve continuous long-term operation without refueling for 10 years assuming a load factor of the core of 44 %. In the present study, arrangement of fuel rods and U enrichment of UO fuel rods were surveyed as design parameters. The U enrichment was determined to be 8.5 wt% to satisfy design requirement. Reactor physics parameters including reactivity coefficients and power distributions were evaluated for the determined core specifications. Reactor physics parameters related to core safety were also analyzed. Results of the safety analysis indicated that the determined core specifications satisfied design conditions. Reactor shutdown performance by dropping the reflector, which was adopted as a passive reactor shutdown system, was analyzed and confirmed it's availability from a viewpoint of reactor physics.

Report on neutronic design calculational methods

; *; *; *

JNC TN8410 2000-011, 185 Pages, 2000/05

JNC-TN8410-2000-011.pdf:4.67MB

This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.

Preparation of next generation set of group cross sections; A Task report to the Japan Nuclear Cycle Development Institute)

*

JNC TJ9400 2000-005, 182 Pages, 2000/03

JNC-TJ9400-2000-005.pdf:4.74MB

The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radial heterogereous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By coparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and O, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. Bycoaparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it ...

Measurement of neutron capture cross sections of Tc-99

*

JNC TJ9400 2000-008, 61 Pages, 2000/02

JNC-TJ9400-2000-008.pdf:2.5MB

For studies on nuclear transmutation of long-lived fission products (LLFPs) in a fast reactor, detailed characteristics of reactor core such as transmutation performance have to be investigated, so accurate neutron cross section data of LLFPs become necessary. Therefore, the keV-neutron capture cross sections of Tc-99, which is one of important LLFPs, were measured in the present study to obtain the accurate data. The measurement was relative to the standard capture cross sections of Au-197. A neutron time-of-flight method was adopted with a ns-pulsed neutron source by a Pelletron accelerator and a large anti-Compton NaI(TI) gamma-ray detector. As a result, the capture cross sections of Tc-99 were obtained with the error of about 5 % in the incident neutlon energy region of 10 to 600 keV. The present data were compared with other experimental data and the evaluated values of JENDL-3.2, and it was found that the evaluations of JENDL-3.2 were 15-20 % smaller than the present measurements.

Preparation of methods to calculate pin-wise intra-subassembly power density distribution of a new in-pile experimental reactor for FBR safety research

Mizuno, Masahiro*; Uto, Nariaki

JNC TN9400 98-007, 147 Pages, 1998/11

JNC-TN9400-98-007.pdf:8.32MB

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 ...

None

PNC TN1410 97-034, 338 Pages, 1997/09

PNC-TN1410-97-034.pdf:6.65MB

no abstracts in English

None

*; *; *

PNC TJ1409 97-011, 25 Pages, 1997/03

PNC-TJ1409-97-011.pdf:0.59MB

None

Development of a standard data base for FBR core nuclear design (VI): JUPITER-II experimental data book

*

PNC TN9450 96-052, 694 Pages, 1996/10

PNC-TN9450-96-052.pdf:45.48MB

The present report compiles the experimental data of JUPITER Phase-II, which was a joint research program between U.S. DOE and PNC of Japan, using the ZPPR facility, which stands for Zero Power Physics Reactor at ANL-Idaho in l982 to l984. The JUPITER-II experiment was a series of critical experiments for conventional radial heterogeneous cores of 650 MWe class LMFBR, including six experimental cores. The nuclear characteristics recorded here include criticality, control rod reactivity, reaction rate distribution, sodium void reactivity, sample reactivity, Doppler reactivity and gamma heating. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The detail of experimental data is thoroughly recorded here so as to re-analyze these experiments in future. In addition, these experimental data are installed in the computer system at OEC for convenience of analytical code input.