Study on the production of ¹⁷⁷Lu for medical purposes  at the Dalat Research Reactor. Part 1. Study on production of ¹⁷⁷Lu at the Dalat Research Reactor

Lutetium-177 (177Lu) radioisotope used for medical purposes has been produced at the Dalat Nuclear Research Reactor during 2012-2014. The product was synthezed by the activation reaction 176Lu (n,γ) 177Lu. The target was Lutetium oxide with an abundance of 176Lu being 2.59% and the irradiation was conducted with a neutron flux of 2.1013 n cm-2 s-1. The activity of the 177Lu product was 39,59 Ci g-1 after 108 hous of the irradiation. The yield of this method was much higher compared to those of the reaction 176Yb(n, γ)177Yb → 177Lu + β- which for the same time could be 15mCi/50mg only. Because of low specific radioactivity the preparation of 177Lu made from the activation of 176Yb cannot be applied for the medical purposes. Additionally, the separation yield of 177Lu from 177Yb is rather low, it is usualy of approximately 70%.

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Characterization of neutron spectrum parameters at irradiation channels for neutron activation analysis after full conversion of the Dalat nuclear research reactor to low enriched uranium fuel

Nuclear Science and Technology ◽

10.53747/jnst.v4i1.216 ◽

2014 ◽

Vol 4 (1) ◽

pp. 70-75

Author(s):

D. Vu C. ◽

Q. Thien T. ◽

V. Doanh H. ◽

D. Quyet P. ◽

T. Anh T.T. ◽

...

Keyword(s):

Neutron Activation Analysis ◽

Neutron Flux ◽

Neutron Spectrum ◽

Thermal Neutron Flux ◽

Research Reactor ◽

Nuclear Research ◽

Fast Neutron Flux ◽

Nuclear Research Reactor ◽

Enriched Uranium ◽

Full Core

In the framework of the program on Russian Research Reactor Fuel Return (RRRFR) and the program on Reduced Enrichment for Research and Test Reactor (RERTR), the full core conversion of the Dalat Nuclear Research Reactor (DNRR) to low enriched uranium (LEU, 19.75% 235U) fuel was performed from November 24, 2011 to January 13, 2012. The reactor is now operated with a working core consisting of 92 WWR-M2 LEU. After the full core conversion, the neutron spectrum parameters which are used in k0-NAA such as thermal neutron flux (fth), fast neutron flux (ffast), f factor, alpha factor (a), and neutron temperature (Tn) have been re-characterized at four different irradiated channels in the core. Based on the experimental results, it can be seen that the thermal neutron flux decreases by 6÷9% whereas fast neutron flux increases by 2÷6%. The neutron spectrum becomes‘harder’ at most of irradiated positions. The obtained neutron spectrum parameters from this research are used to re-establish the procedures for Neutron Activation Analysis (NAA) according to ISO/IEC 17025:2005 standard at NuclearResearch Institute.

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Some Main Results of Commissioning of The Dalat Research Reactor with Low Enriched Fuel

Nuclear Science and Technology ◽

10.53747/jnst.v4i1.212 ◽

2014 ◽

Vol 4 (1) ◽

pp. 36-45

Author(s):

Nhi Dien Nguyen ◽

Ba Vien Luong ◽

Van Lam Pham ◽

Vinh Vinh Le ◽

Ton Nghiem Huynh ◽

...

Keyword(s):

Research Reactor ◽

Nuclear Research ◽

Experimental Results ◽

Design Calculation ◽

Uranium Fuel ◽

Nuclear Research Reactor ◽

Enriched Uranium ◽

Design Calculations ◽

Dalat Research Reactor ◽

Good Agreement

After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements.

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Design of a FPGA-based controller for power and period measurement in the start range of Dalat Nuclear Research Reactor

Nuclear Science and Technology ◽

10.53747/jnst.v10i3.7 ◽

2021 ◽

Vol 10 (3) ◽

pp. 41-48

Author(s):

Vo Van Tai ◽

Nguyen Van Kien ◽

Nguyen Nhi Dien ◽

Trinh Dinh Hai ◽

Le Van Diep

Keyword(s):

Neutron Flux ◽

High Speed ◽

Research Reactor ◽

Full Range ◽

Reactor Core ◽

Nuclear Research ◽

Reactor Power ◽

Processing Unit ◽

Upward Movement ◽

Nuclear Research Reactor

This paper introduces a new controller module based on a high-speed field-programmable gate array (FPGA) and digital signal processing (DSP) using moving average (MA) filters for calculation of the reactor power and period at the start range of the Dalat nuclear research reactor (DNRR). The reactor power is proportional to the neutron flux in the reactor core, and the reactor period is the time that the reactor power changes by a factor of 2.718. In the control and protection system (CPS) of the DNRR, the reactor power and period have been monitored by the 8-bit microprocessor controller named BPM-107R. There are two main functions of the BPM-107R controller including 1) measurement and determination of reactor power and period and 2) generation of warning and emergency protection signals by reactor power or/and by reactor period. Those discrete signals will access to the logical processing unit of the CPS to prohibit the upward movement of control rods or to shut down the reactor. The CPS has three BPM-107R controllers corresponding to three independent neutron flux measurement equipment (NFME) channels working by logic voting “2 out-of 3”. Each NFME channel was designed for detection of neutron flux density in the full range from 1×100 to 1.2×1010 n/cm2 ×s, which is divided into two sub-ranges named start range (SR) and working range (WR). The designed FPGA-based controller module was tested using simulated signals as well as signals from the CPS in comparison with the original controller BPM-107R. The experimental results show that the characteristics and functions of the two controllers are equivalent.

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Simulation of absorbed doses distribution in a polyethylene phantom for BNCT application at the Dalat research reactor

Nuclear Science and Technology ◽

10.53747/jnst.v10i4.9 ◽

2021 ◽

Vol 10 (4) ◽

pp. 01-07

Author(s):

Pham Dang Quyet ◽

Pham Ngoc Son ◽

Nguyen Nhi Dien ◽

Nguyen An Son ◽

Trinh Thi Tu Anh ◽

...

Keyword(s):

Thermal Neutron ◽

Research Reactor ◽

Absorbed Dose ◽

Nuclear Research ◽

Gamma Dose ◽

Central Axis ◽

Maximum Value ◽

Nuclear Research Reactor ◽

Mcnp5 Code ◽

Dalat Research Reactor

In this paper, the distribution of absorbed dose components in a polyethylene phantom for BNCT application at Dalat Nuclear Research Reactor (DNRR) were calculated using the MCNP5 code. The configuration of horizontal neutron channel No.2 of the DNRR, which contains a cylindrical collimator with neutron filters of 20-cm Si and 3-cm Bi, was simulated. The results show that the gamma dose along the central axis of the phantom has the maximum value of 1.82×10-6 Gy at the 0.5-cm depth, and reduces to 9.05×10-7 Gy at the 3-cm depth. The main contribution to gamma dose is due to the interaction of thermal neutron with hydrogen in the phantom via the 1H(n,γ)2H reaction, and its value is much smaller than thermal neutron dose. The total absorbed dose along the central axis of the phantom has the maximum value of 7.87×10-5 Gy at the 0.5-cm depth, and decreases rapidly to 1.52×10-5 Gy at the 3-cm position, and mainly depends on the boron and thermal neutron doses caused by the 10B(n, α)7Li and 14N(n, p)14C reactions, respectively.

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