• Nuclear Engineering and Technology
• /
• 제56권1호
• /
• pp.106-113
• /
• 2024

Understanding the tritium release and retention behavior of candidate tritium breeder materials is crucial for breeder blanket design. Recently, a melt spraying process was developed to prepare Li₄SiO₄ pebbles, which were subsequently subjected to the in-pile tritium production and extraction platform in China Mianyang Research Reactor (CMRR) to investigate their in-situ tritium release behavior and irradiation performance. The results demonstrate that HT is the main tritium release form, and adding hydrogen to the purge gas reduces tritium retention while increasing the HT percent in the purge gas. Post-irradiation experiments reveal that the irradiated pebbles darken in color and their grains swell, but the mechanical properties remain largely unchanged. It is concluded that the tritium residence time of Li₄SiO₄ made by melt spraying method at 467 ℃ is approximately 23.34 h. High-density Li₄SiO₄ pebbles exhibit tritium release at relatively low temperatures (<600 ℃) that is mainly controlled by bulk diffusion. The diffusion coefficient at 525 ℃ and 550 ℃ is 1.19 × 10^-11 cm²/s and 5.34 × 10^-11 cm²/s, respectively, with corresponding tritium residence times of 21.3 hours and 4.7 hours.

• Nuclear Engineering and Technology
• /
• 제49권6호
• /
• pp.1219-1225
• /
• 2017

In this work, a scalable algorithm for model calibration in nuclear engineering applications is presented and tested. The algorithm relies on the construction of surrogate models to replace the original model within the region of interest. These surrogate models can be constructed efficiently via reduced order modeling and subspace analysis. Once constructed, these surrogate models can be used to perform computationally expensive mathematical analyses. This work proposes a surrogate based model calibration algorithm. The proposed algorithm is used to calibrate various neutronics and thermal-hydraulics parameters. The virtual environment for reactor applications-core simulator (VERA-CS) is used to simulate a three-dimensional core depletion problem. The proposed algorithm is then used to construct a reduced order model (a surrogate) which is then used in a Bayesian approach to calibrate the neutronics and thermal-hydraulics parameters. The algorithm is tested and the benefits of data assimilation and calibration are highlighted in an uncertainty quantification study and requantification after the calibration process. Results showed that the proposed algorithm could help to reduce the uncertainty in key reactor attributes based on experimental and operational data.

• Nuclear Engineering and Technology
• /
• 제51권1호
• /
• pp.269-274
• /
• 2019

$^{64}Cu$ is a favorable radionuclide in nuclear medicine applications because of its unique characteristics such as three types of decay (electron capture, ${\beta}^-$ and ${\beta}^+$) and 12.7 h half-life. Production of $^{64}Cu$ by irradiation $^{nat}Cu$ and $^{nat}CuNPs$ in Tehran Research Reactor was investigated. The characteristics of copper nanoparticles were investigated with SEM, TEM and XRD analysis. The cross section of $^{63}Cu(n,{\gamma})^{64}Cu$ reaction was done with TALYS-1.8 code. The activity value of $^{64}Cu$ was calculated with theoretical approach and MCNPX-2.6 code. The results were compared with related experimental results which showed good adaptations between them.

• Nuclear Engineering and Technology
• /
• 제53권11호
• /
• pp.3517-3527
• /
• 2021

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).

• Journal of Radiation Protection and Research
• /
• 제25권3호
• /
• pp.157-159
• /
• 2000

• Nuclear Engineering and Technology
• /
• 제56권8호
• /
• pp.2893-2905
• /
• 2024

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
• /
• pp.771-776
• /
• 1998

TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Vnit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) far all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library. BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The makimum fast neutron nun calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effgctive full power days is 1.784x10$^{18}$ n/$\textrm{cm}^2$. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60cm below the midplane at zero degree.

• Nuclear Engineering and Technology
• /
• 제50권1호
• /
• pp.165-169
• /
• 2018

An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was $7.40{\times}10^5Bq/cm^3$ and at the bay region was $3.39{\times}10^5Bq/cm^3$. At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was $4.14{\times}10^{-1}Bq/cm^3$.

• Nuclear Engineering and Technology
• /
• 제51권8호
• /
• pp.1964-1969
• /
• 2019

The safety and efficiency of nuclear reactors largely depend on the monitoring and control of nuclear radiation. Due to the unique nuclear-physical characteristics, Hf is one of the most promising materials for the manufacturing of the control rods and the emitters of neutron detectors. It is proposed to use the Compton neutron detector with the emitter made of Hf in the In-core Instrumentation System (ICIS) for monitoring the neutron field. The main advantages of such a detector in comparison the conventional β-emission sensors are the possibility of reaching of a higher cumulative radiation dose and the absence of signal delays. The response time of the detection is extremely important when a nuclear reactor is operating near its critical operational parameters. Taking Hf as an example, the general principles for calculating the chains of materials transformation under neutron irradiation are reported. The influence of ^179m1Hf on the Hf composition changing dynamics and the process of transmutants' (Ta, W) generation were determined. The effect of these processes on the absorbing properties of Hf, which inevitably predetermine the lifetime of the detector and its ability to generate a signal, is estimated.

• Journal of Radiation Protection and Research
• /
• 제24권4호
• /
• pp.231-233
• /
• 1999

This report will introduce the epitome about the subject, HANARO (Hi-flux Advanced Neutron Application Reactor designed by KAERI in early 1995) Utilization Education Training Program Development and Operation, which is one of the nuclear research basic expansion businesses executed from 1999. 12. to strengthen the usage of HANARO. This program consists of the basic reactor experiments program for university students who have specialty of nuclear and other engineering, and the special research education program for faculties from universities and researchers from industrial fields. Principle lessons are reactor operation, radioisotope production, neutron activation analysis, neutron radiography, radiation shield (health physics), nuclear fuel combustion measurement by gamma scanning arrangement, and CNS (Compact Nuclear Simulator) and so on.