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Journal Articles

A Study on transmutation of LLFPs using various types of HTGRs

Kora, Kazuki*; Nakaya, Hiroyuki*; Matsuura, Hideaki*; Goto, Minoru; Nakagawa, Shigeaki; Shimakawa, Satoshi*

Nuclear Engineering and Design, 300, p.330 - 338, 2016/04

 Times Cited Count:5 Percentile:38.13(Nuclear Science & Technology)

In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system "GTHTR300", a small modular HTGR plant "HTR50S" and two types of plutonium burner HTGRs "Clean Burn with MA" and "Clean Burn without MA" were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

Journal Articles

Study on transmutation and storage of LLFP using a high-temperature gas-cooled reactor

Kora, Kazuki*; Nakaya, Hiroyuki*; Kubo, Kotaro*; Matsuura, Hideaki*; Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 12 Pages, 2014/09

In this study, the capability of HTGR as LLFP transmuter was evaluated in terms of neutron economy. Considering gas turbine high-temperature reactor with 300 MWe nominal capacity (GTHTR300) as HTGR, transmutations of four types of LLFP nuclide were estimated using Monte Carlo transport code MVP and ORIGEN. In addition, burn-up simulations for whole-core region were carried out using MVP-BURN. It was numerically shown that the neutron fluxes change significantly depending on the arrangement of LLFP in the core. When 15 t of LLFP is placed in an ideal manner, the GTHTR300 can sustain sufficient reactivity for one year while transmuting up to 30 kg per year. Additionally, there are more space available for storing larger amount of LLFP without affecting the reactivity. These results suggest that there is a possibility of using GTHTR300 as both LLFP storage and transmuter.

JAEA Reports

Study on methodology to estimate isotope generation and depletion for core design of HTGR

Fukaya, Yuji; Ueta, Shohei; Goto, Minoru; Shimakawa, Satoshi

JAEA-Research 2013-035, 84 Pages, 2013/12

JAEA-Research-2013-035.pdf:3.22MB

An investigation on methodology to estimate isotope generation and depletion had been performed in order to improve the accuracy for HTGR core design. Solving the burn-up equations, generating effective cross section and employing nuclide data are the technical problems. Especially for the generating effective cross section, the core burn-up calculation has a technological problem in common with point burn-up calculation. Thus, the investigation had also been performed for the core burn-up calculation to develop new code system in the future. As a result, it was found that the cross section with the extended 108 energy groups structure from the SRAC 107 groups structure to 20 MeV and the cross section collapse using the flux obtained by the deterministic code SRAC is proper for the use. In addition, an investigation on the preparation condition for nuclear data for a safety analysis and a fuel design was also performed. As a result, the needs for the nuclear data ware made clear.

Journal Articles

A Study of applicability of JENDL-4.0 to the HTTR criticality analysis

Goto, Minoru; Shimakawa, Satoshi; Tachibana, Yukio

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 6 Pages, 2011/10

In the past, benchmark calculations of criticality approach for the HTTR, which is a Japanese HTGR, were performed by research institutes in several countries, and almost all of the calculations overestimated the excess reactivity. In Japan, the benchmark calculations were performed by JAEA, and were also resulted in overestimation. JAEA improved the calculations by revising the geometric model and replacing the nuclear data library with JENDL-3.3, which was the latest JENDL at that time. However, the overestimation remained and this problem has not been resolved until today. We performed calculations of the HTTR criticality approach with several nuclear data libraries, and found that slight difference in the neutron capture cross section of carbon at thermal energy among the libraries causes significant difference in the $$k_{eff}$$ values. The cross section value of carbon was not concerned in reactor neutronics calculation because of its small value of the order of 1E-3 burn, and consequently the cross section value had not been revised for a long time even in the major nuclear data libraries: JENDL, ENDF/B and JEFF. We thought that the cross section should be revised based on the latest measurement data to improve the accuracy of the HTGR criticality analysis. In May 2010, the latest JENDL (JENDL-4.0) was released by JAEA, and the capture cross section of carbon was revised. JENDL-4.0 yielded 0.4-0.9%$$Delta$$k/k smaller $$k_{eff}$$ values than JENDL-3.3 in the criticality calculations for the HTTR critical approach, and consequently the problem of the overestimation of the excess reactivity in the HTTR benchmark calculation was resolved by replacing the nuclear data libraries with JENDL-4.0.

Journal Articles

JENDL-4.0 benchmark for high temperature gas-cooled reactor, HTTR

Goto, Minoru; Shimakawa, Satoshi; Yasumoto, Takashi*

JAEA-Conf 2011-002, p.11 - 16, 2011/09

In the past, benchmark calculations of the HTTR criticality approach, which is a Japanese HTGR, have been performed by several countries, and almost of the calculations have overestimated its excess reactivity. In Japan, the benchmark calculations were performed by JAEA, and the calculations also resulted in overestimation. JAEA improved this overestimation by revising the problem geometry and replacing nuclear data library with JENDL-3.3, which was the latest JENDL at that time. However, the overestimation remained and this problem had not been resolved until today. We performed the calculation of the HTTR criticality approach with several nuclear data libraries, and found the slightly difference of the capture cross section of carbon at thermal energy among the libraries causes the difference of the $$k$$$$_{eff}$$ values to be not negligible. This cross section value had not been concerned in reactor neutronics calculation because of its small value on the order of 10$$^{-3}$$ burn, and consequently the cross section value had not been revised for a long time even in the major nuclear data libraries: JENDL, ENDF/B and JEFF. We have thought the cross section value should be revised based on the latest measurement data to improve the accuracy of the neutronics calculations for HTTR. On May in 2010, the latest JENDL (JENDL-4) was released by JAEA, and the capture cross section of carbon was revised. Consequently, JENDL-4 yielded 0.4-0.9%$$Delta$$k smaller $$k$$$$_{eff}$$ values than JENDL-3.3 in the calculation for the HTTR critical approach, and then the problem of the overestimation of the excess reactivity in the HTTR benchmark calculation was resolved.

Journal Articles

Core design study of small-sized high temperature reactor for electricity generation

Goto, Minoru; Shimakawa, Satoshi; Terada, Atsuhiko; Shibata, Taiju; Tachibana, Yukio; Kunitomi, Kazuhiko

Proceedings of ASME 2011 Small Modular Reactors Symposium (SMR 2011) (CD-ROM), 5 Pages, 2011/09

The present study challenges the core design of a small-sized reactor for long refueling interval by increasing core size, fuel loading and fuel burn up compared with the High Temperature engineering Test Reactor (HTTR). The core burn-up calculation suggested that approximately 6 years of long refueling interval was found to be reasonably achieved with operational reactor power of 120 MWt.

Journal Articles

Impact of revised thermal neutron capture cross section of carbon stored in JENDL-4.0 on HTTR criticality calculation

Goto, Minoru; Shimakawa, Satoshi; Nakao, Yasuyuki*

Journal of Nuclear Science and Technology, 48(7), p.965 - 969, 2011/07

 Times Cited Count:13 Percentile:23.58(Nuclear Science & Technology)

In the past, benchmark calculations of criticality approach for the HTTR, which is a Japanese HTGR, were performed by research institutes in several countries, and almost all of the calculations overestimated the excess reactivity. In Japan, the benchmark calculations performed by JAEA also resulted in overestimation. JAEA improved the calculations by revising the geometric model and replacing the nuclear data library with JENDL-3.3, which was the latest JENDL at that time. However, the overestimation remained and this problem has not been resolved until today. We performed calculations of the HTTR criticality approach with several nuclear data libraries, and found that slight difference in the capture cross section of carbon at thermal energy among the libraries causes significant difference in the $$k$$$$_{eff}$$ values. The cross section value of carbon was not concerned in reactor neutronics calculation because of its small value of the order of 10$$^{-3}$$ burn, and consequently the cross section value was not revised for a long time even in the major nuclear data libraries: JENDL, ENDF/B and JEFF. We thought that the cross section value should be revised based on the latest measurement data in order to improve the accuracy of the neutronics calculations of the HTTR. In April 2010, the latest JENDL;JENDL-4, was released by JAEA, and the capture cross section of carbon was revised. JENDL-4 yielded 0.4%$$Delta$$$$k$$-0.9%$$Delta$$$$k$$ smaller $$k$$$$_{eff}$$ values than JENDL-3.3 in the calculation of the HTTR critical approach, and consequently the problem of the overestimation of the excess reactivity in the HTTR benchmark calculation was resolved.

Journal Articles

Long-term high-temperature operation in the HTTR, 2; Core physics

Goto, Minoru; Fujimoto, Nozomu; Shimakawa, Satoshi; Tachibana, Yukio; Nishihara, Tetsuo; Iyoku, Tatsuo

Proceedings of 5th International Topical Meeting on High Temperature Reactor Technology (HTR 2010) (CD-ROM), 8 Pages, 2010/10

In the High Temperature Engineering Test Reactor (HTTR), which is a Japanese block-type HTGR, reactivity is controlled by control rods (CRs) and burnable poisons (BPs). The CRs insertion depth into the core should be retained shallow during burnup period, because the large insertion depth leads to significant disturbance of the power distribution, and consequently fuel temperature rises above the limit. Thus, the controllable reactivity with the CRs during operation is small, and then reactivity control through the burnup period largely depends on the BPs. It has not been confirmed an effectiveness of BPs on reactivity control on block-type HTGRs. The HTTR succeeded in long-term high temperature operation, and its burnup reached about 370EFPD. Thereby it became possible to confirm the effectiveness of BPs on reactivity control on the HTTR using its burnup data. We focused on a burnup change in the CRs insertion depth into the core to confirm whether the BPs functioned as designed. Additionally, we compared the change in the CRs insertion depths between analysis results and the experimental data to confirm validity of a whole core burnup calculation with the SRAC/COREBN. As a result, the experimental data showed that although the CRs insertion depth into the core was increased with burnup, it was retained the allowable depth. Meanwhile, the analysis result of the CRs insertion depth was in good agreement with the experimental data.

Journal Articles

Evaluation of required activity of SO$$_{3}$$ decomposition catalyst for iodine-sulfur process

Imai, Yoshiyuki; Kubo, Shinji; Goto, Minoru; Shimakawa, Satoshi; Tachibana, Yukio; Onuki, Kaoru

Proceedings of 5th International Topical Meeting on High Temperature Reactor Technology (HTR 2010) (CD-ROM), 4 Pages, 2010/10

Required performance of SO$$_{3}$$ decomposition catalyst for Iodine-Sulfur process was investigated. Heat transfer area needed for shell and tube type SO$$_{3}$$ decomposer exchanging heat from VHTR was calculated by applying Yagi-Kunii and Zukauskas's equation for filled layer-SiC tube and SiC tube-He gas flow heat transfer respectively and the minimum space velocity for catalyst was 1000 h$$^{-1}$$. To transform minimum space velocity to more universal kinetic rate constant, we introduced forward/reverse SO$$_{3}$$ decomposition equation. To achieve equilibrium SO$$_{3}$$ decomposition ratio above 0.5 MPa, rate constant k$$_{1}$$ should be more than 1.5 s$$^{-1}$$ for SO$$_{3}$$ decomposition catalyst.

Journal Articles

Impact of capture cross-section of carbon on nuclear design for HTGRs

Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki; Tachibana, Yukio

Proceedings of 5th International Topical Meeting on High Temperature Reactor Technology (HTR 2010) (CD-ROM), 6 Pages, 2010/10

Capture cross section of carbon in thermal energy range has been regarded as unimportant in neutronics calculations on general reactor design, because of its infinitesimal value of only 3 mb at 2200 m/s. However, it is not negligible for design works for graphite-rich reactors, such as the High Temperature Gas-cooled Reactors (HTGRs). For the High Temperature Engineering Test Reactor (HTTR) of JAEA, five percent differences in capture cross section of carbon makes 0.24% change in thermal utilization factor of the four factor formula. This impact is for the HTTR with a core configuration of full-loaded core, named the packed core. In this case, change of multiplier factor will be equivalent to a change of thermal utilization factor. The impact of the cross section is dependent on an atomic number ratio of graphite/235-uranimu in reactor core. For more graphite-rich core such as the HTTR with ring core configuration, the five percent change of the cross section value makes a 0.47%$$Delta$$$$k$$ on multiplier factor. From our studies in the HTTR analysis, a value of capture cross section at 2200 m/s has been revised to 3.86 mb in evaluated nuclear data library of JENDL-4. Comparing with the value of JENDL4, the values in other libraries are about 10-15% smaller as 3.36 mb in ENDF/B-VII, 3.36 mb in JEFF-3.1 and 3.53 mb in JENDL-3.3. It was observed that discrepancy of a multiplier factor between former calculation and experiment of the HTTR showed disagreement in the evaluation of the critical approach tests. Monte Carlo calculation results using JENDL3.3 are overestimated about 0.4%$$Delta$$$$k$$ with packed core configuration and 1.0%$$Delta$$$$k$$ with ring core, respectively. In this report, the improvement of excess reactivity calculation for the HTTR with newly JENDL-4 is described.

Journal Articles

Irradiation test of component for radiation-resistant small-sized motor

Nakamichi, Masaru; Ishitsuka, Etsuo; Shimakawa, Satoshi; Kan, Satoshi*

Fusion Engineering and Design, 84(7-11), p.1399 - 1403, 2009/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Irradiation tests of a small-sized motor with radiation resistance

Nakamichi, Masaru; Ishitsuka, Etsuo; Shimakawa, Satoshi; Kan, Satoshi*

Fusion Engineering and Design, 83(7-9), p.1321 - 1325, 2008/12

 Times Cited Count:1 Percentile:88.61(Nuclear Science & Technology)

no abstracts in English

Journal Articles

VHTR Deep Burn applications

Richards, M.*; Venneri, F.*; Shimakawa, Satoshi

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

A key feature of the Very High Temperature Reactor (VHTR) is its use of ceramic, coated-particle fuel that can be irradiated at very high temperatures to very high burnup. In this paper we describe applications for deep-burn VHTRS (DB-VHTRs) using Pu, transuranic (TRU), and mixed oxide (MOX) fuels. One application is efficient and economic disposition of surplus weapons-grade Pu (WPu). A DB-VHTR using WPu fuel can consume about 65% of the WPu and about 90% of the Pu- 239 in a single pass through the core (corresponding to approximately 600 GWt-d/t). Comparable levels of burnup can be achieved using TRU material recovered from light water reactor (LWR) spent fuel. DB-VHTRs can also be deployed with Fast Breeder Reactors (FBRs) in a completely closed fuel cycle to provide both sustainability of nuclear fuel resources and flexible energy outputs, including hydrogen production and other process-heat applications.

JAEA Reports

Examination on small-sized cogeneration HTGR for developing countries

Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; et al.

JAEA-Technology 2008-019, 57 Pages, 2008/03

JAEA-Technology-2008-019.pdf:8.59MB

The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors.

Journal Articles

Neutronics calculations of HTTR with several nuclear data libraries

Goto, Minoru; Nojiri, Naoki; Shimakawa, Satoshi

Journal of Nuclear Science and Technology, 43(10), p.1237 - 1244, 2006/10

 Times Cited Count:4 Percentile:66.5(Nuclear Science & Technology)

Benchmark calculations for several HTTR core conditions were performed with four cross-section sets which were generated from JENDL-3.3, JENDL-3.2, ENDF/B-6.8 and JEFF-3.0 using a continuous energy Monte Carlo code MVP. The core conditions were an annular form core at room temperature and a full core with cylindrical form at room temperature and at full power operation. Study of k$$_{eff}$$ discrepancies caused by difference of the nuclear data libraries and identification of nuclide which gives large effects on the k$$_{eff}$$ discrepancies were carried out. Comparison of the k$$_{eff}$$ between calculations and experiments was also carried out. As the results, for each HTTR core conditions, JENDL-3.3 gives the k$$_{eff}$$ agreement with the experiments within 1.5%$$Delta$$k, JENDL-3.2 gives the k$$_{eff}$$ agreement within 1.7%$$Delta$$k, and ENDF/B-6.8 and JEFF-3.0 give k$$_{eff}$$ agreement within 1.8%$$Delta$$k. There is little k$$_{eff}$$ discrepancy between ENDF/B-6.8 and JEFF-3.0. The $$k_{eff}$$ discrepancy between JENDL-3.3 and JENDL-3.2 is caused by difference of U-235 data and has temperature dependency. The k$$_{eff}$$ discrepancy between JENDL-3.3 and ENDF/B-6.8 or JEFF-3.0 is caused by difference of graphite data.

Journal Articles

Assessment of calculation model for annular core on the HTTR

Nojiri, Naoki; Handa, Yuichi*; Shimakawa, Satoshi; Goto, Minoru; Kaneko, Yoshihiko*

Nippon Genshiryoku Gakkai Wabun Rombunshi, 5(3), p.241 - 250, 2006/09

It was shown from the annular core experiment of the HTTR that the discrepancy of excess reactivity between experiment and analysis reached about 3 % Dk/k at maximum. Sensitivity analysis for the annular core of the HTTR was performed to improve the discrepancy. The SRAC code system was used for the core analysis. As the results of the analysis, it was found clearly that the multiplication factor of the annular core is affected by (1) mesh interval in the core diffusion calculation, (2) mesh structure of graphite region in fuel lattice cell and (3) the Benoist's anisotropic diffusion coefficients. The significantly large discrepancy previously reported was reduced down to about 1 % Dk/k by the revised annular core model.

Journal Articles

Neutron powder diffraction study on the crystal and magnetic structures of BiCoO$$_3$$

Belik, A. A.*; Iikubo, Satoshi; Kodama, Katsuaki; Igawa, Naoki; Shamoto, Shinichi; Niitaka, Seiji*; Azuma, Masaki*; Shimakawa, Yuichi*; Takano, Mikio*; Izumi, Fujio*; et al.

Chemistry of Materials, 18(3), p.798 - 803, 2006/02

 Times Cited Count:230 Percentile:1.42(Chemistry, Physical)

The crystal and magnetic structures of polycrystalline BiCoO$$_3$$ have been determined by the Rietveldmethod from neutron diffraction data measured at temperatures from 5 to 520 K. BiCoO$$_3$$ (space groupP4mm; Z=1; a=3.72937(7) $AA and c=4.72382(15) AA at room temperature; tetragonality c/a=1.267)is isotypic with BaTiO$_3$$ and PbTiO$$_3$$ in the whole temperature range. BiCoO$$_3$$ is an insulator with a Neeltemperature of 470 K. A possible model for antiferromagnetic order is proposed with a propagationvector of k=(1/2, 1/2, 0). In this model, magnetic moments of Co$$^{3+}$$ ions are parallel to the c directionand align antiferromagnetically in the ab plane. The antiferromagnetic ab layers stack ferromagneticallyalong the c axis, forming a C-type antiferromagnetic structure. Refined magnetic moments at 5 and 300K are 3.24(2)$$mu$$$$_B$$ and 2.93(2)$$mu$$$$_B$$, respectively. The structure refinements revealed no deviation fromstoichiometry in BiCoO$$_3$$. BiCoO$$_3$$ decomposed in air above 720 K to give Co$$_3$$O$$_4$$ and sillenite-like Bi$$_{25}$$CoO$$_{39}$$.

JAEA Reports

Rise-to-power test in high temperature engineering test reactor in the high temperature test operation mode; Test progress and summary of test results up to 30MW of reactor thermal power

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tochio, Daisuke; Shimakawa, Satoshi; Nojiri, Naoki; Goto, Minoru; Shibata, Taiju; Ueta, Shohei; et al.

JAERI-Tech 2004-063, 61 Pages, 2004/10

JAERI-Tech-2004-063.pdf:3.14MB

The High Temperature engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30MW and the reactor outlet coolant temperature of 850$$^{circ}$$C/950$$^{circ}$$C. Rise-to-power test in the HTTR was performed from March 31th to May 1st in 2004 as phase 5 test up to 30MW in the high temperature test operation mode. It was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30MW and 950$$^{circ}$$C respectively on April 19th in the single operation mode using only the primary pressurized water cooler. The parallel loaded operation mode using the intermediate heat exchanger and the primary pressurized water cooler was performed from June 2nd and JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test on June 24th from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests were passed successfully in the high temperature test operation mode. Achievement of the reactor-outlet coolant temperature of 950$$^{circ}$$C is the first time in the world. It is possible to extend highly effective power generation with a high-temperature gas turbine and produce hydrogen from water with a high-temperature. This report describes the results of the high-temperature test operation of the HTTR.

Journal Articles

Characteristic test of initial HTTR core

Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

Nuclear Engineering and Design, 233(1-3), p.283 - 290, 2004/10

 Times Cited Count:9 Percentile:44.08(Nuclear Science & Technology)

This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations.

JAEA Reports

Core dynamics analysis of control rod withdrawal test in HTTR (Contract Research)

Takada, Eiji*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu

JAERI-Tech 2004-048, 60 Pages, 2004/06

JAERI-Tech-2004-048.pdf:4.18MB

The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395$$^{circ}$$C and coolant outlet temperature of 850$$^{circ}$$C/950$$^{circ}$$C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. The reactivity insertion test demonstrates that rapid increase of reactor power by withdrawing the control rod is restrained by only the negative reactivity feedback effect without operating the reactor power control system, and the temperature transient of the reactor is slow. The best estimated analyses have been conducted to simulate reactor transients during the reactivity insertion test. A one-point core dynamics approximation with one fuel channel model is applied to this analysis. It was found that the analytical model for core dynamics could simulate the reactor power behavior.

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