Reactor physics experiment on a graphite-moderated core to construct integral experiment database for HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Nuclear Science and Engineering, 197(8), p.2251 - 2257, 2023/08

https://doi.org/10.1080/00295639.2022.2087836

Thermal-neutron capture cross-section measurements of neptunium-237 with graphite thermal column in KUR

Nakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi

Journal of Nuclear Science and Technology, 59(11), p.1388 - 1398, 2022/11

https://doi.org/10.1080/00223131.2022.2058639

The present study selected Np among radioactive nuclides and aimed to measure the thermal-neutron capture cross-section for Np in a well-thermalized neutron field by an activation method. A Np standard solution was used for irradiation samples. A thermal-neutron flux at an irradiation position was measured with neutron flux monitors: Sc, Co, Mo, Ta and Au. The Np sample and flux monitors were irradiated together for 30 minutes in the graphite thermal column equipped with the Kyoto University Research Reactor. The similar irradiation was carried out twice. After the irradiations, the Np samples were quantified using 312-keV gamma ray emitted from Pa in a radiation equilibrium with Np. The reaction rates of Np were obtained from gamma-ray peak net counts given by Np, and then the thermal-neutron capture cross-section of Np was found to be 173.84.4 barn by averaging the results obtained by the two irradiations. The present result was in agreement with the reported data given by a time-of-flight method within the limit of uncertainty.

Preliminary experiment in a graphite-moderated core to avoid full mock-up experiment for the future first commercial HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 9 Pages, 2022/05

Burn-up characteristics and criticality effect of impurities in the graphite structure of a commercial-scale prismatic HTGR

Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo

Nuclear Engineering and Design, 326, p.108 - 113, 2018/01

AA2015-0964.pdf:0.64MB

https://doi.org/10.1016/j.nucengdes.2017.11.003

Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.

Integral test of international reactor dosimetry and fusion file on graphite assembly with DT neutron at JAEA/FNS

Ota, Masayuki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara

Fusion Engineering and Design, 98-99, p.1847 - 1850, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.03.032

International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), has been released from the International Atomic Energy Agency (IAEA) recently. In order to validate and test IRDFF 1.0, IAEA has initiated a new Co-ordinated Research Project (CRP). Under this CRP, we have performed an integral experiment on a graphite pseudo-cylindrical slab assembly with DT neutron source at JAEA/FNS. The graphite assembly of 31.4 cm in equivalent radius and 61 cm in thickness is placed at a distance of about 20 cm from the DT neutron source. A lot of foils for the dosimetry reactions in IRDFF1.0 are inserted into the small spaces between the graphite blocks along the center axis of the assembly. After DT neutron irradiation, reaction rates for the dosimetry reactions are measured by the foil activation technique. This experiment is analyzed by using Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0. The experimental assembly and DT neutron source are modeled precisely in the MCNP calculation. The reaction rates calculated with IRDFF 1.0 as the response functions for the dosimetry reactions are compared with the experimental values. Also the calculations with JENDL Dosimetry File 99 (JENDL/D-99) are performed for comparison. The results calculated with IRDFF 1.0 show good agreement with the experimental results.

Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

JAEA-Technology 2014-038, 51 Pages, 2014/12

JAEA-Technology-2014-038.pdf:3.84MB

The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.

Characterization of F-irradiated graphite surfaces using photon-stimulated desorption spectroscopy

Sekiguchi, Tetsuhiro; Baba, Yuji; Shimoyama, Iwao; Nath, K. G.

Surface and Interface Analysis, 38(4), p.352 - 356, 2006/04

https://doi.org/10.1002/sia.2172

We investigated the orientation nature at the top-most layers of F-irradiated graphite using polarization dependent near-edge X-ray absorption fine structure (NEXAFS) spectroscopy which incorporates partial electron yield (PEY) detection and photon-stimulated ion desorption (PSID) techniques. The fluorine K-edge NEXAFS spectra conducted in PEY mode show no significant dependence on polarization angles. In contrast, NEXAFS spectra recorded in F ion yield mode show enhanced yields at a feature of 689.4 eV assigned as a *(C-F) state relevant to =C-F sites, which depend on polarization angles. The C-F bonds prefer relatively tilting down the surface at the top-most layer, while the C-F bonds are randomly directed at deeper regions. We conclude that the difference in the orientation structures between the top surface and bulk is reflected in the NEXAFS recorded in the two different detection modes. It was also found that H- and F- PSID NEXAFS spectra are helpful in understanding desorption mechanism, thus in analysing NEXAFS data.

Separation of C from irradiated graphite materials, 1; Oxidation behaviors and the changes in pore structure of Q1 and IG-110 graphite due to the air reaction (Joint research)

Fujii, Kimio

JAERI-Tech 2005-048, 108 Pages, 2005/09

JAERI-Tech-2005-048.pdf:25.05MB

The graphite-moderated power reactor was shut down in 1998 and its decommissioning program is being planned. Various graphites are used in the core of magnox-type reactors and HTTR as core-support structural materials and moderating materials of fast neutrons. For the nuclear graphite disposal, it is necessary to determine especially the treatment of long-lived nuclides, such as C which are generated in the graphite components during reactor operation. As a research, which solves the problem of the C concentration, the cooperative research is concluded between JAERI and Japan Nuclear Power Corp. in 1999, and the research for the basic data acquisition has been advanced up to the present. To find the optimum conditions for C reduction, basic data on oxidation reaction and the structure of graphite materials are indispensable. In the present experiment, we measure the air oxidation characteristics in the temperature range 450800C in Quality1 graphite and IG-110 graphite. Changes in pore diameter and pore size distribution due to air oxidation are discussed.

Research and development plan for advanced high temperature gas cooled reactor fuels and graphite components (Contract research)

Sawa, Kazuhiro; Ueta, Shohei; Shibata, Taiju; Sumita, Junya; Ohashi, Jumpei; Tochio, Daisuke

JAERI-Tech 2005-024, 34 Pages, 2005/03

JAERI-Tech-2005-024.pdf:2.15MB

The Very-High-Temperature Reactor (VHTR) is one of the strong candidates for the Generation IV Nuclear Energy System. JAERI has developed Zirconium carbide (ZrC)-coated fuel particle and ZrC coating layer is expected to maintain its intactness under higher temperature and burn-up comparing conventional SiC-coating layer. JAERI carries out (1) ZrC-coating process development by large-scale coater, (2) inspection method development and (3) irradiation test and post irradiation experiment of ZrC coated particles. Also, JAERI carries out reactivity insertion tests to clarify the coating failure mechanism and tries to increase allowable temperature limit in case of reactivity insertion accident. Furthermore, JAERI develops non-destructive evaluation methods for mechanical properties of graphite components by ultrasonic testing and micro-indentation technique. This report describes these research and development plan and results of FY 2004 as a MEXT contact research.

Study on brittle fracture model for multiaxial tensile stress

Hanawa, Satoshi; Ishihara, Masahiro; Motohashi, Yoshinobu*

Zairyo, 54(2), p.201 - 206, 2005/02

https://doi.org/10.2472/jsms.54.201

no abstracts in English