Performance test of HTTR

Nakagawa, Shigeaki; Tachibana, Yukio; Takamatsu, Kuniyoshi; Ueta, Shohei; Hanawa, Satoshi

doi:10.1016/j.nucengdes.2004.08.027

Cited by 13 publications

(7 citation statements)

References 2 publications

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“…Built at the Oarai Research and Development Center, this advanced reactor first reached criticality in November 1998, followed by full-power operation on December 7, 2001 [1]. Designed and built to further establish High Temperature Gas Reactor (HTGR) technologies, the core has been central to many reactor physics experiments [2,3,4,5] and safety studies [6,7]. Additionally, the core configuration and experiments have been extensively benchmarked [8,9,10].…”

Section: Introductionmentioning

confidence: 99%

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

Labouré

Ortensi

Hummel

2018

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The High Temperature Test Reactor (HTTR) is a graphite-moderated thermal reactor with a nominal power output of 30 MWth. The purpose of this work is to prepare a set of neutron cross sections for use in the transient analyses of three Loss of Forced Cooling (LOFC) events planned for this reactor. The cross sections are generated with the Serpent 2 3-D Monte Carlo code to include the axial heterogeneity and the strong axial coupling of this core. The approach yields good agreement of results when compared to previously used 2-D lattice models, which require a fine spatial discretization and, surprisingly, comes at a higher computational cost. In addition, the temperature distribution must be considered to obtain reasonable cross sections and the axial leakage component dominates the spectral effects in the core. Consequently, a temperature model as a function of the fuel and moderator temperatures is adopted, and the core is depleted to 390 effective full-power days (EFPDs). A full tabulation of cross sections is prepared at this burnup point for a variety of fuel and moderator temperatures. Next, the multiphysics reactor application MAMMOTH is used to evaluate the quality of these cross sections using a diffusion solver. As expected, the homogenization error in the cross sections is significant and the Super Homogenization (SPH) correction from MAMMOTH is necessary to preserve key reaction rates. The SPH-corrected MAMMOTH results are in excellent agreement with the Serpent results and reproduce both the reference power profile and temperature coefficients at each tabulation point. This fact confirms the accuracy of the SPH-corrected cross sections and demonstrates the maturity of using Serpent 2 and MAMMOTH for 3-D cross-section generation, even for reactors as complex as the HTTR. Finally, the authors recommend that the cross sections should be parametrized with the local burnup and that a full core burnup calculation with coupled thermal-fluids would provide a better estimate of the initial condition for any subsequent transient analysis. Unfortunately, this was not within the scope of this work, and direct analysis with these tabulations should be considered as a first-order approximation, at best.

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Section: Introductionmentioning

confidence: 99%

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

Labouré

Ortensi

Hummel

2018

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“…The HTTR (1) was built in the Oarai Research & Development Center of the Japan Atomic Energy Agency (JAEA) to establish and enhance technologies for HTGRs (2) (3) . The HTTR is the first gas-cooled reactor in JAPAN, which is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950 °C (4) (5) .…”

Section: Introductionmentioning

confidence: 99%

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

Takamatsu

Nakagawa

Takeda

2008

JPES

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Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

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“…Edificio del Reactor (Saito et al, 1994).................................................... Figura 2-11. Configuración del elemento de combustible (Nakagawa et al, 2004a). .. Figura 2-12.…”

Section: Agradecimientosunclassified

“…Estructuras internas del Reactor (Shiozawa et al, 2004)......................... Figura A-5. Configuración del elemento de combustible (Nakagawa et al, 2004a). .…”

Section: Agradecimientosunclassified

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Desarrollo de una metodología de análisis del riesgo para sistemas de generación de hidrógeno mediante reactores nucleares.

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Performance test of HTTR

Cited by 13 publications

References 2 publications

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

Desarrollo de una metodología de análisis del riesgo para sistemas de generación de hidrógeno mediante reactores nucleares.

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