Neutronics Calculations of HTTR with Several Nuclear Data Libraries

Goto, Minoru; Nojiri, Naoki; Shimakawa, Satoshi

doi:10.1080/18811248.2006.9711217

Cited by 12 publications

(4 citation statements)

References 7 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The results of criticality calculations for graphite moderated reactors are strongly influenced by the neutron induced capture reaction cross section on carbon that is used in the neutronic calculations. This has been demonstrated by the neutronic calculations done for the High Temperature Engineering Test Reactor (HTTR) in Japan, carried out by Goto et al (2006Goto et al ( , 2011. In Goto et al (2011), criticality calculations for different numbers of loaded fuel columns of the HTTR core were performed using different cross section data libraries.…”

Section: Introductionmentioning

confidence: 99%

Review of the natC(n, γ) cross section and criticality calculations of the graphite moderated reactor BR1

Díez¹,

Stankovskiy²,

Malambu³

et al. 2013

Annals of Nuclear Energy

View full text Add to dashboard Cite

A review of the experimental data for nat C(n,y) and 12 C(n,y) was made to identify the origin of the nat C capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.l libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK«CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12 C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the nat C thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.

show abstract

Section: Introductionmentioning

confidence: 99%

Review of the natC(n, γ) cross section and criticality calculations of the graphite moderated reactor BR1

Díez¹,

Stankovskiy²,

Malambu³

et al. 2013

Annals of Nuclear Energy

View full text Add to dashboard Cite

show abstract

“…Fuel compacts are graphite matrix dispersed, and each compact consists of about 13 000 coated fuel particles with 0.92 mm diameter, as shown in Figure 2. The compact fuel dimension is 39 mm in height, 10 mm inner diameter, and 26 mm in outer diameter 21 …”

Section: Design Conceptmentioning

confidence: 99%

“…The compact fuel dimension is 39 mm in height, 10 mm inner diameter, and 26 mm in outer diameter. 21 For 100 MWt power, additional zones are applied to the radial and axial directions. The zone is divided into five parts for the geometrical change in the radial direction, with the lowest and highest fissile content being 7% and 14%, respectively.…”

Section: Design Optimization and Enrichment Zoningmentioning

confidence: 99%

Comparison of neutronic aspects in high‐temperature gas‐cooled reactor using ZrC and SiC Triso particle with 50 and 100 MWt power

Miftasani

Su’ud

Irwanto

et al. 2021

Intl J of Energy Research

View full text Add to dashboard Cite

Summary The use of zirconium carbide (ZrC) as a substitute for silicon carbide (SiC) has been proposed to improve the performance of coated particles for high‐temperature gas‐cooled reactor (HTGR) fuel. Irradiation test on ZrC has proved that it has excellent characteristics for HTGR fuel compared with SiC, such as fission product retention capabilities and better resistance on fission product corrosion. Neutronic analysis on 30 MWt HTGR using ZrC and SiC has been performed in many previous studies. The research presented in this paper was conducted as a further continuation of the aforementioned studies. The present study was directed to analyze tristructural isotropic (TRISO)‐coated particles with ZrC as a substitute for the SiC layer on an HTGR with different power rates. Here, we used high‐temperature test reactor design, an experimental scale prismatic HTGR built and operated in Japan. The power was varied at the range of 50‐100 MWt, while the reactor geometry was modified by applying an additional fuel layer in the axial and radial directions. Neutronic analysis of the reactor was investigated by calculating the k‐eff, k‐inf, power peaking factor, burn‐up level, neutron spectrum, and Doppler effect using ZrC and SiC layers for TRISO‐coated fuel particles. The result shows the coated fuel particle using the ZrC layer has the same performance as SiC with an insignificant difference in the values on neutronic aspects calculation. Neutronic calculations are performed using the SRAC code with the Japanese Evaluated Nuclear Data Library 4.0 nuclear data library.

show abstract

“…The fuel zone contains 182,000 TRISO per fuel rod, and every block contains 33 fuel rods with 5 blocks per fuel column (or assembly). This corresponds to approximately thirty million TRISO particles in every fuel assembly [3,4,5]. Representing this large number of TRISO particles is a heavy computational load.…”

Section: I-introductionmentioning

confidence: 99%

Core Physics Analysis of An Assembly of HTTR Reactor using Homogeneous and heterogeneous model.

Aziz

2023

Arab Journal of Nuclear Sciences and Applications

View full text Add to dashboard Cite

MCNPX computer code based onMonte Carlo method is used to design a computer model for an assembly of high temperature testing reactor (HTTR). Two models are used in the analysis, namely homogeneous and heterogeneous models. The reactor uses TRISO fuel, Graphite moderator and helium coolant. The multiplication factor of the assembly is determined as a function of fuel burnup and operation time. Axial power mapping distributions are evaluated. Time evolution of actinides (U 235 and Pu 239 ) is calculated as a result of fuel burnup. Fuel and moderator temperature coefficient of reactivity are determined as a function of operating temperature. The effect of Helium coolant losses on reactor criticality is evaluated (by assuming reduction of helium density to one percent of its nominal density). Calculations indicated that homogenous model results are in good agreement with heterogeneous models with an average difference of approximately 5 %.This enable homogeneous model to be used in full reactor core simulations. This is much easier and saves modeling and computational time.

show abstract

Neutronics Calculations of HTTR with Several Nuclear Data Libraries

Cited by 12 publications

References 7 publications

Review of the natC(n, γ) cross section and criticality calculations of the graphite moderated reactor BR1

Review of the natC(n, γ) cross section and criticality calculations of the graphite moderated reactor BR1

Comparison of neutronic aspects in high‐temperature gas‐cooled reactor using ZrC and SiC Triso particle with 50 and 100 MWt power

Core Physics Analysis of An Assembly of HTTR Reactor using Homogeneous and heterogeneous model.

Contact Info

Product

Resources

About