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JAEA Reports

Report of summer holiday practical training on 2022

Ishitsuka, Etsuo; Ho, H. Q.; Kitagawa, Kanta*; Fukuda, Takahito*; Ito, Ryo*; Nemoto, Masaya*; Kusunoki, Hayato*; Nomura, Takuro*; Nagase, Sota*; Hashimoto, Haruki*; et al.

JAEA-Technology 2023-013, 19 Pages, 2023/06

JAEA-Technology-2023-013.pdf:1.75MB

Eight people from five universities participated in the 2022 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the feasibility study for nuclear battery, the burn-up analysis of HTTR core, the feasibility study for $^{252}$ Cf production, the analysis of behavior on loss of forced cooling test, and the thermal-hydraulic analysis near reactor pressure vessel. In the questionnaire after this training, there were impressions such as that it was useful as a work experience, that some students found it useful for their own research, and that discussion with other university students was a good experience. These impressions suggest that this training was generally evaluated as good.

JAEA Reports

Calculation of nuclear core parameters for HTTR; Report of summer holiday practical training 2021

Isogawa, Hiroki*; Naoi, Motomasa*; Yamasaki, Seiji*; Ho, H. Q.; Katayama, Kazunari*; Matsuura, Hideaki*; Fujimoto, Nozomu*; Ishitsuka, Etsuo

JAEA-Technology 2022-015, 18 Pages, 2022/07

JAEA-Technology-2022-015.pdf:1.37MB

As a summer holiday practical training 2021, the impact of 10 years long-term shutdown on critical control rod position of the HTTR and the delayed neutron fraction ( $_{rm eff}$ ) of the VHTRC-1 core were investigated using Monte-Carlo MVP code. As a result, a long-term shutdown of 10 years caused the critical control rods of the HTTR to withdraw about 4.00.8 cm compared to 3.9 cm in the experiment. The change in critical control rods position of the HTTR is due to the change of some fission products such as $^{241}$ Pu, $^{241}$ Am, $^{147}$ Pm, $^{147}$ Sm, $^{155}$ Gd. Regarding the $_{rm eff}$ calculation of the VHTRC-1 core, the $_{rm eff}$ value is underestimate of about 10% in comparison with the experiment value.

JAEA Reports

Report of summer holiday practical training 2020; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 3

Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.

JAEA-Technology 2021-016, 16 Pages, 2021/09

JAEA-Technology-2021-016.pdf:1.8MB

As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (183 layers) fuel blocks with 20% enrichment of $^{235}$ U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.

JAEA Reports

Mesh effect around burnable poison rod of cell model for HTTR fuel block

Fujimoto, Nozomu*; Fukuda, Kodai*; Honda, Yuki*; Tochio, Daisuke; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo

JAEA-Technology 2021-008, 23 Pages, 2021/06

JAEA-Technology-2021-008.pdf:2.62MB

The effect of mesh division around the burnable poison rod on the burnup calculation of the HTTR core was investigated using the SRAC code system. As a result, the mesh division inside the burnable poison rod does not have a large effect on the burnup calculation, and the effective multiplication factor is closer to the measured value than the conventional calculation by dividing the graphite region around the burnable poison rod into a mesh. It became clear that the mesh division of the graphite region around the burnable poison rod is important for more appropriately evaluating the burnup behavior of the HTTR core..

JAEA Reports

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the $^{235}$ U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with $^{235}$ U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

JAEA Reports

Study on control rod model in HTTR core analysis

Nagasumi, Satoru; Matsunaka, Kazuaki*; Fujimoto, Nozomu*; Ishii, Toshiaki; Ishitsuka, Etsuo

JAEA-Technology 2020-003, 13 Pages, 2020/05

JAEA-Technology-2020-003.pdf:1.5MB

The influence of the control rod model on the nuclear characteristics of the HTTR has been evaluated, by creating detailed control rod model, in which geometric shape was close to that of the actual control rod structure, in MVP code. According to refinement of the control rod model, the critical control rod position was 11 mm lower than that of the conventional model, and this was close to the measured value of 1775 mm. The reactivity absorbed by the shock absorber located at the tip of the control rod was 0.2%k/k, and this was 14 mm difference at the critical control rod position. Considering the effect of refinement of the control rod and the effect of the shock absorber, the correction amount for the analysis value in SRAC code due to the shape effect of the control rod, is -0.05%k/k in reactivity, and -3 mm in the critical control rod position at low temperature criticality.

Journal Articles

Development of fuel temperature calculation code for HTGRs

Inaba, Yoshitomo; Nishihara, Tetsuo

Annals of Nuclear Energy, 101, p.383 - 389, 2017/03

https://doi.org/10.1016/j.anucene.2016.11.006

Times Cited Count：7 Percentile：56.89(Nuclear Science & Technology)

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.

JAEA Reports

Calculation by PHITS code for recoil tritium release rate from beryllium under neutron irradiation (Joint research)

Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2016-022, 35 Pages, 2016/10

JAEA-Technology-2016-022.pdf:3.73MB

As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, the calculation methods by PHITS code is studied to evaluate the recoil tritium release rate from beryllium core components. Calculations using neutron and triton sources were compared, and it is clear that the tritium release rates in both cases show similar values. However, the calculation speed for the triton source cases is two orders faster than that for the neutron source case. It is also clear that the calculation up to history number per unit volume of 210 $^{4}$ (cm $^{-3}$ ) is necessary to determine the recoil tritium release rate of two effective digits precision. Furthermore, the relationship between the beryllium shape and recoil tritium release rate using the triton sources was studied. Recoil tritium release rate showed linear relation to the surface area per volume of beryllium, and the recoil tritium release rate showed about half of the conventional equation value.

Journal Articles

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo

Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08

https://doi.org/10.1080/00223131.2015.1096849

Times Cited Count：1 Percentile：10.71(Nuclear Science & Technology)

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.

Journal Articles

Introduction to modern nodal method and discontinuity factor

Okumura, Keisuke

Nihon Genshiryoku Gakkai Dai-36-Kai Robutsuri Kaki Semina Tekisuto, p.81 - 102, 2004/08

The modern node method which uses a discontinuous factor has come to be widely used recently in the reactor core analyses of commercial light water reactors. The basic theory, numerical computation technique and examples of calculation results are explained for biginners of the modern nodal method.

JAEA Reports

Study on residual radioactive inventory estimation in reactor decommissioning program (Contract research)

Sukegawa, Takenori; Hatakeyama, Mutsuo; Yanagihara, Satoshi

JAERI-Tech 2001-058, 81 Pages, 2001/09

JAERI-Tech-2001-058.pdf:5.98MB

In general, neutron transport and activation calculation codes are used for residual radioactive inventory estimation; however, it is essential to verify calculations by measurement results because of geometrical complexity of the reactor and so on. The comparison between measured and calculated radioactivity in the JPDR core components showed a relatively good agreement (factor of 2), and it was cleared that water content and weight ratio of steel bars to concrete materials significantly influenced the neutron flux distribution in the biological shield (factor of 2-10 error). The measured radioactivity inside of the reactor pressure vessel wall and at the inner part of the biological shield was compared in detail with the calculations to verify the methodology applied to calculations of radioisotope production. Then it was found that the radioactive inventory could be estimated accurately with combination of calculations and measurement of radioactivity in samples and dose rate distribution for planning of dismantling activities.

Journal Articles