Burnup calculation with versatile reactor analysis code system MARBLE2 (interactive execution demo)

Yokoyama, Kenji

Nihon Genshiryoku Gakkai Dai-51-Kai Robutsuri Kaki Semina Tekisuto "Nensho Keisan No Kiso To Jissen", p.95 - 135, 2019/08

The burnup calculation function included in the versatile reactor analysis code system system MARBLE2 is introduced by an interactive execution demo. Although the main purpose of MARBLE2 is to analyze nuclear characteristics of fast reactors, the users can use it while assembling small functions according to purpose. Therefore, it can be applied other purposes than the nuclear characteristic analysis of fast reactors. In order to realize such usage, MARBLE is developed by using an object-oriented scripting language Python. As the Python implementation is short and easy to understand, the burnup function of MARBLE is explained by showing several examples of the implementation. In addition, an example of constructing a simple burnup calculation system using MARBLE is introduced.

Analysis of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya

Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02

https://doi.org/10.1080/00223131.2017.1384705

The criticality safety of the fuel debris in Fukushima Daiichi Nuclear Power Plant is one of the most important issues and the adoption of the burnup credit is desired for the criticality analysis. The assay data of used nuclear fuel irradiated in 2F2 is evaluated to validate SWAT4.0 for BWR fuel burnup problem. The calculation results revealed that number density of many heavy nuclides and FPs showed good agreement with the experimental data except for U, Np, Pu and Sm isotopes. The cause of the difference is assumption of the initial number density and void ratio and overestimation of the capture cross section of Np. The C/E-1 values do not depend on the types of fuel rods (UO or UO-GdO) and it is similar to that for the PWR fuel. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of the BWR fuel and it has sufficient accuracy to be adopted in the burnup credit evaluation of the fuel debris.

Development of three-dimensional diffusion and burn-up code HIZER for Monju core management

Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro

JAEA-Technology 2014-043, 36 Pages, 2015/02

JAEA-Technology-2014-043.pdf:8.94MB

The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.

Nuclide composition benchmark data set for verifying burnup codes on spent light water reactor fuels

Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kono, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki*

Nuclear Technology, 137(2), p.1 - 16, 2002/02

no abstracts in English

Development of continuous energy Monte Carlo burn-up calculation code MVP-BURN

Okumura, Keisuke; Nakakawa, Masayuki; Kaneko, Kunio*; *

JAERI-Conf 2000-018, p.31 - 41, 2001/01

https://jopss.jaea.go.jp/pdfdata/JAERI-Conf-2000-018.pdf

Burnup calculation codes based on the conventional deterministic approach often encounter difficult problems because of the constraints on the geometry description, limit of approximation on the effective resonance cross-sections, failing of the diffusion approximation due to extremely strong anisotropic or heterogenity. They are, for example, the prediction of burn characteristics of plutonium spot, core design of ultra-small reactors, analysis of the sample material in an irradiation capsule of the research rector. To deal with these problems any time, a burn-up calculation code (MVP-BURN) was developed by using a continuous energy Monte Carlo code MVP. MVP-BURN was validated by comparison with the results of deterministic codes in the international benchmark problems, and by comparison with the measured values of the spent fuel composition irradiated in a commercial reactor.

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.

Burnup calculation with estimated neutron spectrum of JMTR irradiation field; Development of the burnup calculation method for fuel pre-irradiated in the JMTR

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

Integrated burnup calculation code system SWAT

Suyama, Kenya; *; *

JAERI-Data/Code 97-047, 128 Pages, 1997/11

JAERI-Data-Code-97-047.pdf:3.06MB

no abstracts in English

Nuclear and thermal calculations on a hybrid system combining a proton accelerator and a fission reactor core; Study of energy conversion system

;

PNC TN9410 97-029, 39 Pages, 1997/03

PNC-TN9410-97-029.pdf:2.58MB

At present, a high power CW(Continuous Wave) electron linac accelerator is under development in PNC as a part of transmutation study by accelerators. Last year we performed nuclear and thermal calculation on a hybrid reactor system combining the electron linac and a subcritical core with TRU fuel as one of applied uses. It is well known that the hybrid reactor system can also use a proton accelerator instead of the electron one. The nuclear and thermal calculation was performed on the system using the proton accelerator in this year. Comparison of the extinguished quantity of TRU fuel was performed between calculation results of both systems. In the proton linac hybrid system, a proton beam accelerated from the linac is injected into a target located in the center of the subcritical core to produce neutrons by spallation reactions. The neutrons enter the surrounding subcritical core to extinguish the TRU. The result of calculation showed that the extinguished quantity of TRU was about 10 kg where the proton beam power of 100 MW injected into the subcritical core system of keff = 0.95 over 1 year. The extinguished quantity of TRU was about 100 times as large as that in the case studied last year on the electron linac hybrid system.

Neutronic analysis of JRR-4 with low enriched uranium silicon dispersion type fuel of 3.8g/cm uranium density

Nakano, Yoshihiro

JAERI-Tech 95-002, 63 Pages, 1995/02

JAERI-Tech-95-002.pdf:2.03MB

no abstracts in English