Design Analyses for Full Core Conversion of  The Dalat Nuclear Research Reactor

The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.

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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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Modeling of the Dalat Nuclear Research Reactor (DNRR) with the Serpent 2 Monte Carlo code

Nuclear Science and Technology ◽

10.53747/jnst.v9i3.41 ◽

2019 ◽

Vol 9 (3) ◽

pp. 21-29

Author(s):

Cuong Nguyen Kien ◽

Dung Nguyen Thi ◽

Phu Tran Viet ◽

Tiep Nguyen Huu ◽

Ha Pham Nhu Viet

Keyword(s):

Monte Carlo ◽

Cross Sections ◽

Research Reactor ◽

Model Development ◽

Nuclear Research ◽

Monte Carlo Code ◽

Reactor Kinetics ◽

Nuclear Research Reactor ◽

Enriched Uranium ◽

Comparative Results

This paper presents a model development of the Dalat Nuclear Research Reactor (DNRR) loaded with low enriched uranium (LEU) fuel using the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code. The DNRR Serpent 2 model was verified through comparing with the MCNP6 criticality calculations under different reactor conditions. The parameters to be compared include the effective neutron multiplication factor, radial and axial powerdistributions, and thermal neutron flux distributions. The comparative results generally show a good agreement between Serpent 2 and MCNP6 and thus indicate that the DNRR Serpent 2 model can be used for further calculations of the DNRR.

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Interim Storage of the Dalat Nuclear Research Reactor: Radiation Safety Analysis

Science and Technology of Nuclear Installations ◽

10.1155/2020/7327045 ◽

2020 ◽

Vol 2020 ◽

pp. 1-10

Author(s):

Kien-Cuong Nguyen ◽

Vinh-Vinh Le ◽

Ton-Nghiem Huynh ◽

Ba-Vien Luong ◽

Nhi-Dien Nguyen ◽

...

Keyword(s):

Research Reactor ◽

Radiation Safety ◽

Nuclear Research ◽

Severe Case ◽

Safety Analysis ◽

Water Levels ◽

Spent Fuel ◽

Nuclear Research Reactor ◽

Enriched Uranium ◽

Interim Storage

Radiation safety analysis of a new interim storage of the Dalat Nuclear Research Reactor (DNRR) for keeping spent high enriched uranium (HEU) fuel bundles during the core conversion to low enriched uranium (LEU) fuel had been performed and presented. The photon source and decay heat of the spent HEU fuel bundles were calculated using the ORIGEN2.1 code. Gamma dose rates of the spent fuel interim storage were evaluated using the MCNP5 code with various scenarios of water levels in the reactor tank and cooling time. The radiation safety analysis shows that the retention of 106 spent HEU fuel bundles at the interim storage together with a core of 92 LEU fuel bundles meets the requirements of radiation safety. The results indicate that in the most severe case, i.e., the complete loss of water in the reactor tank, the operators still can access the reactor hall to mitigate the accident within a limited time. Particularly, in the control room, the dose rate of about 1.4 μ Sv / h is small enough for people to work normally.

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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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Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

Science and Technology of Nuclear Installations ◽

10.1155/2021/6673162 ◽

2021 ◽

Vol 2021 ◽

pp. 1-10

Author(s):

Kien-Cuong Nguyen ◽

Vinh-Vinh Le ◽

Ton-Nghiem Huynh ◽

Ba-Vien Luong ◽

Nhi-Dien Nguyen

Keyword(s):

Research Reactor ◽

Nuclear Research ◽

Nucleate Boiling ◽

Fuel Cladding ◽

Hydraulic Analysis ◽

The Core ◽

Nuclear Research Reactor ◽

Calculation Results ◽

Enriched Uranium ◽

Thermal Hydraulic Analysis

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

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Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code

Nuclear Science and Technology ◽

10.53747/jnst.v8i1.78 ◽

2021 ◽

Vol 8 (1) ◽

pp. 10-16

Author(s):

Nguyen Thanh Vinh Ho ◽

Vinh Vinh Le ◽

Nhi Dien Nguyen ◽

Kien Cuong Nguyen ◽

Ton Nghiem Huynh ◽

...

Keyword(s):

Power Distribution ◽

Nuclear Physics ◽

Research Reactor ◽

Reactor Core ◽

Calculated Data ◽

Thermal Hydraulics ◽

Fuel Type ◽

Fuel Assemblies ◽

Enriched Uranium ◽

New Research

VVR-KN is one of the low-enriched fuel types to be considered for a new research reactor (RR) of a Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam. This fuel type was qualified by a lead test carried out with three fuel assemblies (FAs) in 6-MWt WWR-K research reactor at the Institute of Nuclear Physics, Kazakhstan. VVR-KN fuel was then used for conversion of the WWR-K reactor core from highly-enriched to low-enriched uranium fuel and the reactor was successfully commissioned in September 2016. PLTEMP is a thermal-hydraulic code with plate and coaxial tube models that seems to be suitable for VVR-KN fuel type. Before using PLTEMP code for thermal-hydraulics analysis of the new RR, a calculation for code validation was performed based on the data of the VVR-KN fuel lead test. First, MCNP5 code was used to calculate the power distribution of WWR-K reactor core with lead test fuel assemblies (LTAs) at the core center. Then, thermal-hydraulics parameters of the LTAs were obtained by using PLTEMP code together with calculated data of the power distribution and the lead test conditions. A comparison between the analytic results and the lead test data was made to confirm the suitability of PLTEMP code for thermal-hydraulics analysis of VVR-KN fuel under forced convection and downward flow conditions.

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Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

Science and Technology of Nuclear Installations ◽

10.1155/2017/2615409 ◽

2017 ◽

Vol 2017 ◽

pp. 1-10 ◽

Cited By ~ 6

Author(s):

Giang Phan ◽

Hoai-Nam Tran ◽

Kien-Cuong Nguyen ◽

Viet-Phu Tran ◽

Van-Khanh Hoang ◽

...

Keyword(s):

Research Reactor ◽

Nuclear Research ◽

Maximum Deviation ◽

Control Rod ◽

The Core ◽

Nuclear Research Reactor ◽

The Difference ◽

Enriched Uranium ◽

Analysis Models ◽

Control Rods

Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.

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